Sample records for nuclear safety center

As an integral part of DOE`s International NuclearSafetyCenter (INSC) at Argonne National Laboratory, the INSC Database has been established to provide an interactively accessible information resource for the world`s nuclear facilities and to promote free and open exchange of nuclearsafety information among nations. The INSC Database is a comprehensive resource database aimed at a scope and level of detail suitable for safety analysis and risk evaluation for the world`s nuclear power plants and facilities. It also provides an electronic forum for international collaborative safety research for the Department of Energy and its international partners. The database is intended to provide plant design information, material properties, computational tools, and results of safety analysis. Initial emphasis in data gathering is given to Soviet-designed reactors in Russia, the former Soviet Union, and Eastern Europe. The implementation is performed under the Oracle database management system, and the World Wide Web is used to serve as the access path for remote users. An interface between the Oracle database and the Web server is established through a custom designed Web-Oracle gateway which is used mainly to perform queries on the stored data in the database tables.

The NuclearSafety Information Center (NSIC) serves as a focal point for the collection, analysis and dissemination of information related to safety problems encountered in the design, analysis, and operation of nuclear facilities. (Author/AB)

This paper describes the operations of the NuclearSafety Analysis Center. Established soon after an accident at the Three Mile Island nuclear power plant near Harrisburg, Pennsylvania, its efforts were initially directed towards a detailed analysis of the accident. Continuing functions include: (1) the analysis of generic nuclearsafety issues,…

The Russian Federation Ministry for Atomic Energy (MINATOM) and the US Department of Energy (USDOE) have formed International NuclearSafetyCenters to collaborate on nuclearsafety research. USDOE established the US Center (ISINSC) at Argonne National Laboratory (ANL) in October 1995. MINATOM established the Russian Center (RINSC) at the Research and Development Institute of Power Engineering (RDIPE) in Moscow in July 1996. In April 1998 the Russian center became a semi-independent, autonomous organization under MINATOM. The goals of the center are to: Cooperate in the development of technologies associated with nuclearsafety in nuclear power engineering; Be international centers for the collection of information important for safety and technical improvements in nuclear power engineering; and Maintain a base for fundamental knowledge needed to design nuclear reactors. The strategic approach is being used to accomplish these goals is for the two centers to work together to use the resources and the talents of the scientists associated with the US Center and the Russian Center to do collaborative research to improve the safety of Russian-designed nuclear reactors. The two centers started conducting joint research and development projects in January 1997. Since that time the following ten joint projects have been initiated: INSC databases--web server and computing center; Coupled codes--Neutronic and thermal-hydraulic; Severe accident management for Soviet-designed reactors; Transient management and advanced control; Survey of relevant nuclearsafety research facilities in the Russian Federation; Computer code validation for transient analysis of VVER and RBMK reactors; Advanced structural analysis; Development of a nuclearsafety research and development plan for MINATOM; Properties and applications of heavy liquid metal coolants; and Material properties measurement and assessment. Currently, there is activity in eight of these projects. Details on each of these

This report is a detailed manual of the information specialist's duties at the NuclearSafety Information Center. Information specialists scan the literature for documents to be reviewed, procure the documents (books, journal articles, reports, etc.), keep the document location records, and return the documents to the plant library or other…

The Radiation Safety Information Computational Center (RSICC) was established in 1963 to collect and disseminate computational nuclear technology in the form of radiation transport, shielding and safety software and corresponding nuclear cross sections. Approximately 1700 nuclear software and data packages are in the RSICC collection, and the majority are applicable to reactor dosimetry.

The Radiation Safety Information Computational Center (RSICC) was established in 1963 to collect and disseminate computational nuclear technology in the form of radiation transport, shielding and safety software and corresponding nuclear cross sections. Approximately 1700 nuclear software and data packages are in the RSICC collection, and the majority are applicable to reactor dosimetry.

The cooperating Nuclear Reaction Data Centers are involved in the compilation and exchange of nuclear reaction data for incident neutrons, charged particles and photons. Individual centers may also have services in other areas, e.g., evaluated data, nuclear structure and decay data, reactor physics, nuclearsafety; some of this information may also be exchanged between interested centers. 20 refs., 1 tab.

The main purpose of this report is to help novice reviewers accelerate their apprenticeship at the NuclearSafety Information Center, a computerized information service sponsored by the U.S. Atomic Energy Commission. Guidelines for reviewers are presented in Part 1; Part 2 contains guidelines for the novice editor. The goal of the reviewers and…

National Academy of Sciences - National Research Council, Washington, DC.

This report covers the general issues involved in nuclearsafety research and points out the areas needing detailed consideration. Topics included are: (1) "Principles of NuclearSafety Research" (examining who should fund, who should conduct, and who should set the agenda for nuclearsafety research); (2) "Elements of a Future Agenda for Nuclear…

This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclearsafety analysis capabilities established and maintained at the NSCR.

Nuclear regulation and safety are discussed from the standpoint of a hypothetical country that is in the process of introducing a nuclear power industry and setting up a regulatory system. The national policy is assumed to be in favor of nuclear power. The regulators will have responsibility for economic, reliable electric production as well as for safety. Reactor safety is divided into three parts: shut it down, keep it covered, take out the afterheat. Emergency plans also have to be provided. Ways of keeping the core covered with water are discussed. (DLC)

Nuclear explosives by their design and intended use require collocation of high explosives and fissile material. The design agencies are responsible for designing safety into the nuclear explosive and processes involving the nuclear explosive. The methodology for ensuring safety consists of independent review processes that include the national laboratories, Operations Offices, Headquarters, and responsible Area Offices and operating contractors with expertise in nuclear explosive safety. A NES Study is an evaluation of the adequacy of positive measures to minimize the possibility of an inadvertent or deliberate unauthorized nuclear detonation, high explosive detonation or deflagration, fire, or fissile material dispersal from the pit. The Nuclear Explosive Safety Study Group (NESSG) evaluates nuclear explosive operations against the Nuclear Explosive Safety Standards specified in DOE O 452.2 using systematic evaluation techniques. These Safety Standards must be satisfied for nuclear explosive operations.

The Advanced Research Workshop (ARW) on Nuclear Materials Safety held June 8-10, 1998, in St. Petersburg, Russia, was attended by 27 Russian experts from 14 different Russian organizations, seven European experts from six different organizations, and 14 U.S. experts from seven different organizations. The ARW was conducted at the State Education Center (SEC), a former Minatom nuclear training center in St. Petersburg. Thirty-three technical presentations were made using simultaneous translations. These presentations are reprinted in this volume as a formal ARW Proceedings in the NATO Science Series. The representative technical papers contained here cover nuclear material safety topics on the storage and disposition of excess plutonium and high enriched uranium (HEU) fissile materials, including vitrification, mixed oxide (MOX) fuel fabrication, plutonium ceramics, reprocessing, geologic disposal, transportation, and Russian regulatory processes. This ARW completed discussions by experts of the nuclear materials safety topics that were not covered in the previous, companion ARW on Nuclear Materials Safety held in Amarillo, Texas, in March 1997. These two workshops, when viewed together as a set, have addressed most nuclear material aspects of the storage and disposition operations required for excess HEU and plutonium. As a result, specific experts in nuclear materials safety have been identified, know each other from their participation in t he two ARW interactions, and have developed a partial consensus and dialogue on the most urgent nuclear materials safety topics to be addressed in a formal bilateral program on t he subject. A strong basis now exists for maintaining and developing a continuing dialogue between Russian, European, and U.S. experts in nuclear materials safety that will improve the safety of future nuclear materials operations in all the countries involved because of t he positive synergistic effects of focusing these diverse backgrounds of

A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

Powerplant systems and procedures that ensure the day-to-day health and safety of people in and around the plant is referred to as operational safety. This safety is the result of careful planning, good engineering and design, strict licensing and regulation, and environmental monitoring. Procedures that assure operational safety at nuclear…

Taking a fresh look at nuclearsafety regulations, this study finds that the mandate and organization of the Nuclear Regulatory Commission (NRC) militate against its making sound decisions. The author criticizes failures to make hard decisions on societal risk, to clarify responsibility, and to implement cost-effective safety measures. Among his recommendations are reorganization of the NRC under a single authoritative administrator, separation of technical issues from social ones, and reform of the Price-Anderson Act. The author concludes that the worst eventuality would be to continue the current state of indecision. 161 references, 6 figures, 4 tables.

The U.S. Nuclear Regulatory Commission (NRC) was formed in 1975 to regulate the various commercial and institutional uses of nuclear energy, including nuclear power plants. The agency succeeded the Atomic Energy Commission, which previously had responsibility for both developing and regulating nuclear activities. Federal research and development work for all energy sources, as well as nuclear weapons production, is now conducted by the U.S. Department of Energy. Under its responsibility to protect public health and safety, the NRC has three principal regulatory functions: (1) establish standards and regulations, (2) issue licenses for nuclear facilities and users of nuclear materials, and (3) inspect facilities and users of nuclear materials to ensure compliance with the requirements. These regulatory functions relate to both nuclear power plants and to other uses of nuclear materials - like nuclear medicine programs at hospitals, academic activities at educational institutions, research work, and such industrial applications as gauges and testing equipment. The NRC places a high priority on keeping the public informed of its work. The agency recognizes the interest of citizens in what it does through such activities as maintaining public document rooms across the country and holding public hearings, public meetings in local areas, and discussions with individuals and organizations.

Steam-heated autoclaves are used in gaseous diffusion uranium enrichment plants to heat large cylinders of UF{sub 6}. Nuclear criticality safety for these autoclaves is evaluated. To enhance criticality safety, systems are incorporated into the design of autoclaves to limit the amount of water present. These safety systems also increase the likelihood that any UF{sub 6} inadvertently released from a cylinder into an autoclave is not released to the environment. Up to 140 pounds of water can be held up in large autoclaves. This mass of water is sufficient to support a nuclear criticality when optimally combined with 125 pounds of UF{sub 6} enriched to 5 percent U{sup 235}. However, water in autoclaves is widely dispersed as condensed droplets and vapor, and is extremely unlikely to form a critical configuration with released UF{sub 6}.

The rise and fall of the nuclear power industry in the United States is a well-documented story with enough socio-technological conflict to fill dozens of scholarly, and not so scholarly, books. Whatever the reasons for the situation we are now in, and no matter how we apportion the blame, the ultimate choice of whether to use nuclear power in this country is made by the utilities and by the public. Their choices are, finally, based on some form of risk-benefit analysis. Such analysis is done in well-documented and apparently logical form by the utilities and in a rather more inchoate but not necessarily less accurate form by the public. Nuclear power has failed in the United States because both the real and perceived risks outweigh the potential benefits. The national decision not to rely upon nuclear power in its present form is not an irrational one. A wide ranging public balancing of risk and benefit requires a classification of risk which is clear and believable for the public to be able to assess the risks associated with given technological structures. The qualitative four-level safety ladder provides such a framework. Nuclear reactors have been designed which fit clearly and demonstrably into each of the possible qualitative safety levels. Surprisingly, it appears that safer may also mean cheaper. The intellectual and technical prerequisites are in hand for an important national decision. Deployment of a qualitatively different second generation of nuclear reactors can have important benefits for the United States. Surprisingly, it may well be the "nuclear establishment" itself, with enormous investments of money and pride in the existing nuclear systems, that rejects second generation reactors. It may be that we will not have a second generation of reactors until the first generation of nuclear engineers and nuclear power advocates has retired. PMID:3340728

This paper shows the general design conditions about ergonomics and safety for control centers in the petrochemical process industry. Some of the topics include guidelines for the optimized workstation design, control room layout, building layout, and lighting, acoustical and environmental design. Also takes into account the safety parameters in the control rooms and centers design. The conditions and parameters shown in this paper come from the standards and global advances on this topic on the most recent publications. And also the work was supplemented by field visits of our team to the control center operations in a petrochemical company, and technical literature search efforts. This guideline will be useful to increase the productivity and improve the working conditions at the control rooms. PMID:22317199

The unique requirements and contraints of space nuclear systems require careful consideration in the development of a safety policy. The NuclearSafety Policy Working Group (NSPWG) for the Space Exploration Initiative has proposed a hierarchical approach with safety policy at the top of the hierarchy. This policy allows safety requirements to be tailored to specific applications while still providing reassurance to regulators and the general public that the necessary measures have been taken to assure safe application of space nuclear systems. The safety policy used by the NSPWG is recommended for all space nuclear programs and missions.

This compilation of notes is presented as a source reference for the criticality safety course. At the completion of this training course, the attendee will: be able to define terms commonly used in nuclear criticality safety; be able to appreciate the fundamentals of nuclear criticality safety; be able to identify factors which affect nuclear criticality safety; be able to identify examples of criticality controls as used as Los Alamos; be able to identify examples of circumstances present during criticality accidents; have participated in conducting two critical experiments; be asked to complete a critique of the nuclear criticality safety training course.

The unique requirements and contraints of space nuclear systems require careful consideration in the development of a safety policy. The NuclearSafety Policy Working Group (NSPWG) for the Space Exploration Initiative has proposed a hierarchical approach with safety policy at the top of the hierarchy. This policy allows safety requirements to be tailored to specific applications while still providing reassurance to regulators and the general public that the necessary measures have been taken to assure safe application of space nuclear systems. The safety policy used by the NSPWG is recommended for all space nuclear programs and missions.

Safety in nuclear power plants (NPPs) in India is a very important topic and it is necessary to dissipate correct information to all the readers and the public at large. In this article, I have briefly described how the safety in our NPPs is maintained. Safety is accorded overriding priority in all the activities. NPPs in India are not only safe but are also well regulated, have proper radiological protection of workers and the public, regular surveillance, dosimetry, approved standard operating and maintenance procedures, a well-defined waste management methodology, proper well documented and periodically rehearsed emergency preparedness and disaster management plans. The NPPs have occupational health policies covering periodic medical examinations, dosimetry and bioassay and are backed-up by fully equipped Personnel Decontamination Centers manned by doctors qualified in Occupational and Industrial Health. All the operating plants are ISO 14001 and IS 18001 certified plants. The Nuclear Power Corporation of India Limited today has 17 operating plants and five plants under construction, and our scientists and engineers are fully geared to take up many more in order to meet the national requirements. PMID:20040970

Safety in nuclear power plants (NPPs) in India is a very important topic and it is necessary to dissipate correct information to all the readers and the public at large. In this article, I have briefly described how the safety in our NPPs is maintained. Safety is accorded overriding priority in all the activities. NPPs in India are not only safe but are also well regulated, have proper radiological protection of workers and the public, regular surveillance, dosimetry, approved standard operating and maintenance procedures, a well-defined waste management methodology, proper well documented and periodically rehearsed emergency preparedness and disaster management plans. The NPPs have occupational health policies covering periodic medical examinations, dosimetry and bioassay and are backed-up by fully equipped Personnel Decontamination Centers manned by doctors qualified in Occupational and Industrial Health. All the operating plants are ISO 14001 and IS 18001 certified plants. The Nuclear Power Corporation of India Limited today has 17 operating plants and five plants under construction, and our scientists and engineers are fully geared to take up many more in order to meet the national requirements. PMID:20040970

This compilation of notes is presented as a source reference for the criticality safety course. It represents the contributions of many people, particularly Tom McLaughlin, the course's primary instructor. At the completion of this training course, the attendee will: be able to define terms commonly used in nuclear criticality safety; be able to appreciate the fundamentals of nuclear criticality safety; be able to identify factors which affect nuclear criticality safety; be able to identify examples of criticality controls as used at Los Alamos; be able to identify examples of circumstances present during criticality accidents; be able to identify examples of computer codes used by the nuclear criticality safety specialist; be able to identify examples of safety consciousness required in nuclear criticality safety.

This compilation of notes is presented as a source reference for the criticality safety course. It represents the contributions of many people, particularly Tom McLaughlin, the course`s primary instructor. At the completion of this training course, the attendee will: be able to define terms commonly used in nuclear criticality safety; be able to appreciate the fundamentals of nuclear criticality safety; be able to identify factors which affect nuclear criticality safety; be able to identify examples of criticality controls as used at Los Alamos; be able to identify examples of circumstances present during criticality accidents; be able to identify examples of computer codes used by the nuclear criticality safety specialist; be able to identify examples of safety consciousness required in nuclear criticality safety.

The US Department of Energy has undertaken an initiative to improve the quality of software used to design and operate their nuclear facilities across the United States. One aspect of this initiative is to revise or create new directives and guides associated with quality practices for the safety software in its nuclear facilities. Safety software includes the safety structures, systems, and components software and firmware, support software and design and analysis software used to ensure the safety of the facility. DOE nuclear facilities are unique when compared to commercial nuclear or other industrial activities in terms of the types and quantities of hazards that must be controlled to protect workers, public and the environment. Because of these differences, DOE must develop an approach to software quality assurance that ensures appropriate risk mitigation by developing a framework of requirements that accomplishes the following goals: {sm_bullet} Ensures the software processes developed to address nuclearsafety in design, operation, construction and maintenance of its facilities are safe {sm_bullet} Considers the larger system that uses the software and its impacts {sm_bullet} Ensures that the software failures do not create unsafe conditions Software designers for nuclear systems and processes must reduce risks in software applications by incorporating processes that recognize, detect, and mitigate software failure in safety related systems. It must also ensure that fail safe modes and component testing are incorporated into software design. For nuclear facilities, the consideration of risk is not necessarily sufficient to ensure safety. Systematic evaluation, independent verification and system safety analysis must be considered for software design, implementation, and operation. The software industry primarily uses risk analysis to determine the appropriate level of rigor applied to software practices. This risk-based approach distinguishes safety

Since the beginning of this century, the so-called 3Ss (NuclearSafety, Nuclear Security and Safeguards) have become major regulatory areas for peaceful uses of nuclear energy. In order to rationalize the allocation of regulatory resources, interrelationship of the 3Ss should be investigated. From the viewpoint of the number of the parties concerned in regulation, nuclear security is peculiar with having “aggressors” as the third party. From the viewpoint of final goal of regulation, nuclear security in general and safeguards share the goal of preventing non-peaceful uses of nuclear energy, though the goal of anti-sabotage within nuclear security is rather similar to nuclearsafety. As often recognized, safeguards are representative of various policy tools for nuclear non-proliferation. Strictly speaking, it is not safeguards as a policy tool but nuclear non-proliferation as a policy purpose that should be parallel to other policy purposes (nuclearsafety and nuclear security). That suggests “SSN” which stands for Safety, Security and Non-proliferation is a better abbreviation rather than 3Ss. Safeguards as a policy tool should be enumerated along with nuclearsafety regulation, nuclear security measures and trade controls on nuclear-related items. Trade controls have been playing an important role for nuclear non-proliferation. These policy tools can be called “SSST” in which Trade controls are also emphasized along with Safety regulation, Security measures and Safeguards.

The Pantex Plant, located in the Texas panhandle near Amarillo, is a major Department of Energy (DOE) participant in maintaining the safety of the nation's nuclear weapons resources and protecting the employees, public, and environment. With more than 168,000 person-years of operations involving nuclear materials, explosives, and hazardous chemicals, Pantex has maintained a notable safety record. This article overviews the nuclear weapon activities at Pantex and describes their safety culture. PMID:11045518

The established concept of ensuring safety for nuclear power sources is presented; the influence of severe accidents on nuclear power development is considered, including the accident at a Japan NPP in 2011, as well as the role of state regulation of nuclearsafety.

Nuclearsafety is reviewed in relation to safety in the community as a whole. A method is proposed which points to an optimum expenditure on nuclearsafety measures as opposed to the present open-ended situation. At this optimum point the cost of saving extra lives in the nuclear field is equal to the cost of saving extra lives in other activities in the community. The method requires that the present level of safety be estimated, and this is done by relating the work of Rasmussen, Farmer and Beattie; and the recent German study to the actual record of accidents. The analysis indicates that present expenditures on reactor safety are far in excess of the optimum. An even more striking conclusion is reached when the possible effect of the wealth-generated by the nuclear industry on the general safety of the community is considered. The application of the theme to the Pickering Nuclear Generating Station is developed.

The paper presents the history of safety devices used in nuclear weapons from the early days of separables to the latest advancements in MicroElectroMechanical Systems (MEMS). Although the paper focuses on devices, the principles of Enhanced Nuclear Detonation Safety implementation will also be presented.

The most important concern in the design, construction and operation of nuclear powerplants is safety. Nuclear power is one of the major contributors to the nation's supply of electricity; therefore, it is important to assure its safe use. Each different type of powerplant has special design features and systems to protect health and safety. One…

The importance of safety culture in the safe and reliable operation of nuclear organizations is not a new concept. The greatest barriers to this area of research are twofold: (1) the definition and criteria of safety culture for a nuclear organization and (2) the measurement of those attributes in an objective and systematic fashion. This paper will discuss a proposed resolution of those barriers as demonstrated by the collection of data across nuclear and non-nuclear facilities over a two year period.

Under the likely scenario in which public support for nuclear energy remains low and fossil fuels continue to be abundant and cheap, government supported nuclear research centers must adapt their missions to ensure that they tackle problems of current significance. It will be critical to be multidisciplinary, to generate economic value, and to apply nuclear competencies to current problems. Addressing problems in nuclearsafety, D and D, nuclear waste management, nonproliferation, isotope production are a few examples of current needs in the nuclear arena. Argonne's original mission, to develop nuclear reactor technology, was a critical need for the U.S. in 1946. It would be wise to recognize that this mission was a special instance of a more general one--to apply unique human and physical capital to long term, high risk technology development in response to society's needs. International collaboration will enhance the collective chances for success as the world moves into the 21st century.

This compilation of notes is presented as a source reference for the criticality safety course. It represents the contributions of many people, particularly Tom McLaughlin, the course's primary instructor. At the completion of this training course, the attendee will: (1) be able to define terms commonly used in nuclear criticality safety; (2) be able to appreciate the fundamentals of nuclear criticality safety; (3) be able to identify factors which affect nuclear criticality safety; (4) be able to identify examples of criticality controls as used at Los Alamos; (5) be able to identify examples of circumstances present during criticality accidents; (6) be able to identify examples of safety consciousness required in nuclear criticality safety.

This compilation of notes is presented as a source reference for the criticality safety course. It represents the contributions of many people, particularly Tom McLaughlin, the course`s primary instructor. At the completion of this training course, the attendee will: (1) be able to define terms commonly used in nuclear criticality safety; (2) be able to appreciate the fundamentals of nuclear criticality safety; (3) be able to identify factors which affect nuclear criticality safety; (4) be able to identify examples of criticality controls as used at Los Alamos; (5) be able to identify examples of circumstances present during criticality accidents; (6) be able to identify examples of safety consciousness required in nuclear criticality safety.

An interagency NuclearSafety Working Group (NSPWG) was chartered to recommend nuclearsafety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program. These recommendations, which are contained in this report, should facilitate the implementation of mission planning and conceptual design studies. The NSPWG has recommended a top-level policy to provide the guiding principles for the development and implementation of the SEI nuclear propulsion safety program. In addition, the NSPWG has reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. These recommendations should be useful for the development of the program's top-level requirements for safety functions (referred to as Safety Functional Requirements). The safety requirements and guidelines address the following topics: reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, safeguards, risk/reliability, operational safety, ground testing, and other considerations.

The Defense Nuclear Facilities Safety Board (Board) is an independent federal agency established by Congress in 1988 to provide nuclearsafety oversight of activities at U.S. Department of Energy (DOE) defense nuclear facilities. The activities under the Board's jurisdiction include the design, construction, startup, operation, and decommissioning of defense nuclear facilities at DOE sites. This paper reviews the Board's safety oversight of decommissioning activities at DOE sites, identifies the safety problems observed, and discusses Board initiatives to improve the safety of decommissioning activities at DOE sites. The decommissioning of former defense nuclear facilities has reduced the risk of radioactive material contamination and exposure to the public and site workers. In general, efforts to perform decommissioning work at DOE defense nuclear sites have been successful, and contractors performing decommissioning work have a good safety record. Decommissioning activities have recently been completed at sites identified for closure, including the Rocky Flats Environmental Technology Site, the Fernald Closure Project, and the Miamisburg Closure Project (the Mound site). The Rocky Flats and Fernald sites, which produced plutonium parts and uranium materials for defense needs (respectively), have been turned into wildlife refuges. The Mound site, which performed R and D activities on nuclear materials, has been converted into an industrial and technology park called the Mound Advanced Technology Center. The DOE Office of Legacy Management is responsible for the long term stewardship of these former EM sites. The Board has reviewed many decommissioning activities, and noted that there are valuable lessons learned that can benefit both DOE and the contractor. As part of its ongoing safety oversight responsibilities, the Board and its staff will continue to review the safety of DOE and contractor decommissioning activities at DOE defense nuclear sites.

In August 1945, U.S. Navy Captain William Parsons served as the weaponeer aboard the Enola Gay for the mission to Hiroshima (Shelton 1988). In view of the fact that four B-29s had crashed and burned on takeoff from Tinian the night before, Captain Parsons made the decision to arm the gun-type weapon after takeoff for safety reasons (15 kilotons of TNT equivalent). Although he had no control over the success of the takeoff, he could prevent the possibility of a nuclear detonation on Tinian by controlling what we now call the nuclear explosive. As head of the Ordnance Division at Los Alamos and a former gunnery officer, Captain Parsons clearly understood the role of safety in his work. The advent of the pre-assembled implosion weapon where the high explosive and nuclear materials are always in an intimate configuration meant that nuclear explosive safety became a reality at a certain point in development and production not just at the time of delivery by the military. This is the only industry where nuclear materials are intentionally put in contact with high explosives. The agency of the U.S. Government responsible for development and production of U.S. nuclear weapons is the Department of Energy (DOE) (and its predecessor agencies). This paper will be limited to nuclear explosive safety as it is currently practiced within the DOE nuclear weapons

In nuclear criticality safety applications, a number of important uncertainties have to be addressed to establish the required criticality safety margin of a nuclear system. One source of these uncertainties is the basic nuclear data used to calculate the effective multiplication factor of the system. Before criticality safety calculations are performed, the bias and uncertainties of the codes and cross sections that are used must be determined. Cross-section data are measured, evaluated, and tested prior to their inclusion in nuclear data libraries. Traditionally, nuclear data evaluations are performed to support the analysis and design of thermal and fast reactors. The neutron spectra characteristic of the thermal and fast systems used for data testing are predominantly in the low- and high-energy ranges, with a relatively minor influence from the intermediate-energy range. In the area of nuclear criticality safety, nuclear systems involving spent fuel elements from reactors can lead to situations very different from those most commonly found in reactor analysis and design. These systems are not limited to thermal or fast neutron spectra and may have their most significant influence from the intermediate energy range. This requires extending the range of applicability of the nuclear data evaluation beyond thermal and fast systems. The aim here is to focus on the evaluated nuclear data pertaining to applications in nuclear criticality safety.

This document is the text of a speech given by Pierre Y. Tanguy (Electricite de France) at the 22nd Water Reactor Safety Meeting held in Bethesda, MD in 1994. He describes the EDF nuclear program in broad terms and proceeds to discuss operational safety results with EDF plants. The speaker also outlines actions to enhance safety planned for the future, and he briefly mentions French cooperation with the Chinese on the Daya Bay project.

Nuclear Wallet Cards present properties for ground and isomeric states of all known nuclides. Properties given are: spin and parity assignments, nuclear mass excesses, half-life, isotopic abundances, and decay modes. Appendices contain properties of elements, fundamental constants and other useful information. Nuclear Wallet Cards booklet is published by the National Nuclear Data Center and its electronic (current) version is periodically updated. The Nuclear Wallet Cards by Dr. Jagdish K. Tuli, presently in its 8th edition, is distributed in print as well as in PDA-adaptable Palm Pilot format; the data table as an ASCII file is available upon request. [Taken from http://www.nndc.bnl.gov/wallet/

The DOE Nuclear Criticality Safety Program (NCSP) provides the technical infrastructure maintenance for those technologies applied in the evaluation and performance of safe fissionable-material operations in the DOE complex. These technologies include an Analytical Methods element for neutron transport as well as the development of sensitivity/uncertainty methods, the performance of Critical Experiments, evaluation and qualification of experiments as Benchmarks, and a comprehensive Nuclear Data program coordinated by the NCSP Nuclear Data Advisory Group (NDAG).The NDAG gathers and evaluates differential and integral nuclear data, identifies deficiencies, and recommends priorities on meeting DOE criticality safety needs to the NCSP Criticality Safety Support Group (CSSG). Then the NDAG identifies the required resources and unique capabilities for meeting these needs, not only for performing measurements but also for data evaluation with nuclear model codes as well as for data processing for criticality safety applications. The NDAG coordinates effort with the leadership of the National Nuclear Data Center, the Cross Section Evaluation Working Group (CSEWG), and the Working Party on International Evaluation Cooperation (WPEC) of the OECD/NEA Nuclear Science Committee. The overall objective is to expedite the issuance of new data and methods to the DOE criticality safety user. This paper describes these activities in detail, with examples based upon special studies being performed in support of criticality safety for a variety of DOE operations.

Traditionally, nuclear data evaluations have been performed in support of the analysis and design of thermal and fast reactors. In general, the neutron spectra characteristic of the thermal and fast systems used for data testing are predominantly in the low- and high-energy range with a relatively small influence from the intermediate-energy range. In the area of nuclear criticality safety, nuclear systems arising from applications involving fissionable materials outside reactors can lead to situations very different from those most commonly found in reactor analysis and design. These systems are not limited to thermal or fast and may have significant influence from the intermediate energy range. The extension of the range of applicability of the nuclear data evaluation beyond thermal and fast systems is therefore needed to cover problems found in nuclear criticality safety. Before criticality safety calculations are performed, the bias and uncertainties of the codes and cross sections that are used must be determined. The most common sources of uncertainties, in general, are the calculational methodologies and the uncertainties related to the nuclear data, such as the microscopic cross sections, entering into the calculational procedure. The aim here is to focus on the evaluated nuclear data pertaining to applications in nuclear criticality safety.

The results of a study to identify potential hazards arising from nuclear reactor power systems for use on the lunar and Martian surfaces, related safety issues, and resolutions of such issues by system design changes, operating procedures, and other means are presented. All safety aspects of nuclear reactor power systems from prelaunch ground handling to eventual disposal were examined consistent with the level of detail for SP-100 reactor design at the 1988 System Design Review and for launch vehicle and space transport vehicle designs and mission descriptions as defined in the 90-day Space Exploration Initiative (SEI) study. Information from previous aerospace nuclearsafety studies was used where appropriate. Safety requirements for the SP-100 space nuclear reactor system were compiled. Mission profiles were defined with emphasis on activities after low earth orbit insertion. Accident scenarios were then qualitatively defined for each mission phase. Safety issues were identified for all mission phases with the aid of simplified event trees. Safety issue resolution approaches of the SP-100 program were compiled. Resolution approaches for those safety issues not covered by the SP-100 program were identified. Additionally, the resolution approaches of the SP-100 program were examined in light of the moon and Mars missions.

As the United States embarks on a new era of nuclear arms control, the tools for treaty verification must be accurate and reliable, and must work at stand-off distances. The National Center for Nuclear Security, or NCNS, at the Nevada National Security Site, is poised to become the proving ground for these technologies. The center is a unique test bed for non-proliferation and arms control treaty verification technologies. The NNSS is an ideal location for these kinds of activities because of its multiple environments; its cadre of experienced nuclear personnel, and the artifacts of atmospheric and underground nuclear weapons explosions. The NCNS will provide future treaty negotiators with solid data on verification and inspection regimes and a realistic environment in which future treaty verification specialists can be trained. Work on warhead monitoring at the NCNS will also support future arms reduction treaties.

As the United States embarks on a new era of nuclear arms control, the tools for treaty verification must be accurate and reliable, and must work at stand-off distances. The National Center for Nuclear Security, or NCNS, at the Nevada National Security Site, is poised to become the proving ground for these technologies. The center is a unique test bed for non-proliferation and arms control treaty verification technologies. The NNSS is an ideal location for these kinds of activities because of its multiple environments; its cadre of experienced nuclear personnel, and the artifacts of atmospheric and underground nuclear weapons explosions. The NCNS will provide future treaty negotiators with solid data on verification and inspection regimes and a realistic environment in which future treaty verification specialists can be trained. Work on warhead monitoring at the NCNS will also support future arms reduction treaties.

There has been increased public interest in the potential effects of nuclear powerplant accidents since the Soviet reactor accident at Chernobyl. People have begun to look for more information about the amount of radioactivity that might be released into the environment as a result of such an accident. When this issue is discussed by people…

The concept of safety functions is discussed. Ten critical safety functions and the multiple success paths available for accomplishing them are described. Use of the safety function concept in the development of emergency procedures, operator training, and control-room displays provides a systematic approach and a hierarchy of protection that an operator can use to mitigate the consequences of an event. The safety function concept can also be applied to the design and analysis of nuclear plant systems and to the evaluation of past expierience.

Nuclear propulsion is necessary for successful Mars exploration to enhance crew safety and reduce mission costs. Safety concerns are considered by some to be an implements to the use of nuclear thermal rockets for these missions. Therefore, an assessment was made of the various types of possible accident conditions that might occur and whether design or operational solutions exist. With the previous work on the NERVA nuclear rocket, most of the issues have been addressed in some detail. Thus, a large data base exist to use in an agreement. The assessment includes evaluating both ground, launch, space operations and disposal conditions. The conclusion is that design and operational solutions do exist for the safe use of nuclear thermal rockets and that both the environment and crews be protected against harmful radiation. Further, it is concluded that the use of nuclear thermal propulsion will reduce the radiation and mission risks to the Mars crews.

The Criticality Safety Information Resource Center (CSIRC) at Los Alamos National Laboratory (LANL) is a program jointly funded by the U.S. Department of Energy (DOE) and the U.S. Nuclear Regulatory Commission (NRC) in conjunction with the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 97-2. The goal of CSIRC is to preserve primary criticality safety documentation from U.S. critical experimental sites and to make this information available for the benefit of the technical community. Progress in archiving criticality safety primary documents at the LANL archives as well as efforts to make this information available to researchers are discussed. The CSIRC project has a natural linkage to the International Criticality Safety Benchmark Evaluation Project (ICSBEP). This paper raises the possibility that the CSIRC project will evolve in a fashion similar to the ICSBEP. Exploring the implications of linking the CSIRC to the international criticality safety community is the motivation for this paper.

Nuclear power and propulsion has been considered for space applications since the 1950s. Between 1955 and 1972 the US built and tested over twenty nuclear reactors / rocket engines in the Rover/NERVA programs1. The Aerojet Corporation was the prime contractor for the NERVA program. Modern changes in environmental laws present challenges for the redevelopment of the nuclear rocket. Recent advances in fuel fabrication and testing options indicate that a nuclear rocket with a fuel composition that is significantly different from those of the NERVA project can be engineered; this may be needed to ensure public support and compliance with safety requirements. The Center for Space Nuclear Research (CSNR) is pursuing a number of technologies, modeling and testing processes to further the development of safe, practical and affordable nuclear thermal propulsion systems.

The safety criteria that have been specified for modern nuclear weapons are very demanding. The majority of the weapons in the current stockpile will have to be modified to meet them, unless they are retired. Moreover, for some weapons we still lack necessary data to perform credible safety analyses. Although plutonium dispersal is a much less threatening danger than a sizable nuclear yield, it is nevertheless a potentially serious hazard, particularly if the plutonium is aerosolized in a chemical detonation. The panel recommended the following actions: (1) equip all stockpiled weapons with Enhanced Nuclear Detonation Safety, and build all aircraft-launched bombs and cruise missiles with insensitive high explosive and fire-resistant cores; (2) began an immediate review of the acceptability of retaining missile systems without IHE, fire-resistant cores, or 1.3 class propellant in close proximity to the warheads; (3) continue safety studies and allocate necessary resources to weapons and military laboratories; (4) affirm enhanced safety as the top priority goal of the US nuclear weapons program, and design all future weapons to be as safe as practically achievable, consistent with reasonable military requirements.

The policy of the United States for all US nuclear power sources in space is to ensure that the probability of release of radioactive material and the amounts released are such that an undue risk is not presented, considering the benefits of the mission. The objective of this document is to provide safety criteria which a mission/reactor designer can use to help ensure that the design is acceptable from a radiological safety standpoint. These criteria encompass mission design, reactor design, and radiological impact limitation requirements for safety, and the documentation required. They do not address terrestrial operations, occupational safety or system reliability except where the systems are important for radiological safety. Specific safety specifications based on these criteria shall also be generated and made part of contractual requirements.

The National Nuclear Data Center has provided remote access to its databases and other resources since 1986. This year we have completed the modernization of our databases and Web site. Resources available from our Web site will be summarized and some of the major improvements described in more detail.

A project definition study was conducted for the development, implementation and operation of a National Traffic Safety Documentation Center. Included in this final comprehensive report are: (1) the results of nationwide surveys of users and sources of traffic safety information; (2) a review of relevant information technology in terms of the…

The purpose of this report is to identify and document the nuclearsafety design requirements that are specific to structures, systems, and components (SSCs) of the repository that are important to safety (ITS) during the preclosure period and to support the preclosure safety analysis and the license application for the high-level radioactive waste (HLW) repository at Yucca Mountain, Nevada. The scope of this report includes the assignment of nuclearsafety design requirements to SSCs that are ITS and does not include the assignment of design requirements to SSCs or natural or engineered barriers that are important to waste isolation (ITWI). These requirements are used as input for the design of the SSCs that are ITS such that the preclosure performance objectives of 10 CFR 63.111(b) [DIRS 173273] are met. The natural or engineered barriers that are important to meeting the postclosure performance objectives of 10 CFR 63.113(b) and (c) [DIRS 173273] are identified as ITWI. Although a structure, system, or component (SSC) that is ITS may also be ITWI, this report is only concerned with providing the nuclearsafety requirements for SSCs that are ITS to prevent or mitigate event sequences during the repository preclosure period.

The purpose of this report is to identify and document the nuclearsafety design requirements that are specific to structures, systems, and components (SSCs) of the repository that are important to safety (ITS) during the preclosure period and to support the preclosure safety analysis and the license application for the high-level radioactive waste (HLW) repository at Yucca Mountain, Nevada. The scope of this report includes the assignment of nuclearsafety design requirements to SSCs that are ITS and does not include the assignment of design requirements to SSCs or natural or engineered barriers that are important to waste isolation (ITWI). These requirements are used as input for the design of the SSCs that are ITS such that the preclosure performance objectives of 10 CFR 63.111 [DIRS 156605] are met. The natural or engineered barriers that are important to meeting the postclosure performance objectives of 10 CFR 63.113 [DIRS 156605] are identified as ITWI. Although a structure, system, or component (SSC) that is ITS may also be ITWI, this report is only concerned with providing the nuclearsafety requirements for SSCs that are ITS to prevent or mitigate event sequences during the repository preclosure period.

In the years 1976 to 1982 officialdom intensified the safety regulations in German nuclear power plants out of all proportion, without actually bringing about a recognizable plus in safety or indeed a greater acceptance by the public of the peaceful use of nuclear energy. Although the risk to employees of nuclear power plants and to the population living in their vicinity is substantially smaller than the dangers of modern civilization, the general public still regards with concern the consequences of radioactive exposure and the hazards to later generations from long-life radionuclides. The task for the coming years must be to maintain the safety standard now attained, while simultaneously reducing those exaggerated individual requirements in order to establish a balance in safety precautions. Additionally, a proposal put forward by Sir Walter Marshall, Chairman of the CEGB, should be pursued, i.e., to put the presumed risks of nuclear energy into their correct perspective in the public eye using comprehensible comparisons such as the risks from active or passive smoking. This cannot be accomplished by quoting abstract statistics.

TOPAZ-2 Nuclear Power System (NPS) safety philosophy is based on the requirement that the reactor shall not be critical during all kinds of operations prior to its start-up on the safe orbit (except for physical start-up). Potentially dangerous operation were analyzed and both computational and experimental studies were carried out.

The safety evaluation report (SER) for Galileo was prepared by the Interagency NuclearSafety Review Panel (INSRP) coordinators in accordance with Presidential directive/National Security Council memorandum 25. The INSRP consists of three coordinators appointed by their respective agencies, the Department of Defense, the Department of Energy (DOE), and the National Aeronautics and Space Administration (NASA). These individuals are independent of the program being evaluated and depend on independent experts drawn from the national technical community to serve on the five INSRP subpanels. The Galileo SER is based on input provided by the NASA Galileo Program Office, review and assessment of the final safety analysis report prepared by the Office of Special Applications of the DOE under a memorandum of understanding between NASA and the DOE, as well as other related data and analyses. The SER was prepared for use by the agencies and the Office of Science and Technology Policy, Executive Office of the Present for use in their launch decision-making process. Although more than 20 nuclear-powered space missions have been previously reviewed via the INSRP process, the Galileo review constituted the first review of a nuclear power source associated with launch aboard the Space Transportation System.

Topics discussed in this report include: nuclear utility organization; before the Florida Public Service Commission in re: St. Lucie Unit No. 2 cost recovery; nuclear reliability improvement and safety operations; nuclear utility management; training of nuclear facility personnel; US experience in key areas of nuclearsafety; the US Nuclear Regulatory Commission - function and process; regulatory considerations of the risk of nuclear power plants; overview of the processes of reliability and risk management; management significance of risk analysis; international and domestic institutional issues for peaceful nuclear uses; the role of the Institute of Nuclear Power Operations (INPO); and nuclearsafety activities of the International Atomic Energy Agency (IAEA).

Full text of publication follows: ensuring safety during all stages of facility life cycle is a widely recognised responsibility of the operators, implemented under the supervision of the regulatory body and other competent authorities. As the majority of the facilities worldwide are still in operation or shutdown, there is no substantial experience in decommissioning and evaluation of safety during decommissioning in majority of Member States. The need for cooperation and exchange of experience and good practices on ensuring and evaluating safety of decommissioning was one of the outcomes of the Berlin conference in 2002. On this basis during the last three years IAEA initiated a number of international projects that can assist countries, in particular small countries with limited resources. The main IAEA international projects addressing safety during decommissioning are: (i) DeSa Project on Evaluation and Demonstration of Safety during Decommissioning; (ii) R{sup 2}D{sup 2}P project on Research Reactors Decommissioning Demonstration Project; and (iii) Project on Evaluation and Decommissioning of Former Facilities that used Radioactive Material in Iraq. This paper focuses on the DeSa Project activities on (i) development of a harmonised methodology for safety assessment for decommissioning; (ii) development of a procedure for review of safety assessments; (iii) development of recommendations on application of the graded approach to the performance and review of safety assessments; and (iv) application of the methodology and procedure to the selected real facilities with different complexities and hazard potentials (a nuclear power plant, a research reactor and a nuclear laboratory). The paper also outlines the DeSa Project outcomes and planned follow-up activities. It also summarises the main objectives and activities of the Iraq Project and introduces the R{sup 2}D{sup 2} Project, which is a subject of a complementary paper.

The National Aeronautics and Space Administration (NASA) launched the Jet Propulsion Laboratory's (JPL) two Voyager spacecraft to Jupiter in 1977, each using three radioisotope thermoelectric generators (RTGs) supplied by the Department of Energy (DOE) for onboard electric power. In 1986 NASA will launch JPL's Galileo spacecraft to Jupiter equipped with two DOE supplied RTGs of an improved design. NASA and JPL are also responsible for obtaining a single RTG of this type from DOE and supplying it to the European Space Agency as part of its participation in the International Solar Polar Mission. As a result of these missions, JPL has been deeply involved in space nuclearsafety as a user. This paper will give a brief review of the user contributions by JPL - and NASA in general - to the nuclearsafety processes and relate them to the overall nuclearsafety program necessary for the launch of an RTG. The two major safety areas requiring user support are the ground operations involving RTGs at the launch site and the failure modes and probabilities associated with launch accidents.

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

A qualitative identification of the steps required to assure the incorporation of radiological system safety principles and objectives into all phases of a manned space base program are presented. Specific areas of emphasis include: (1) radiological program management, (2) nuclear system safety plan implementation, (3) impact on program, and (4) summary of the key operation and design guidelines and requirements. The plan clearly indicates the necessity of considering and implementing radiological system safety recommendations as early as possible in the development cycle to assure maximum safety and minimize the impact on design and mission plans.

Summarized is data from a recent study on nuclear reactor safety completed by Norman C. Rasmussen and others. Non-nuclear events are about 10,000 times more likely to produce large accidents than nuclear plants. (RH)

The Office of the National Coordinator for Health Information Technology is expected to oversee creation of a Health Information Technology (HIT) SafetyCenter. While its functions are still being defined, the center is envisioned as a public-private entity focusing on promotion of HIT related patient safety. We propose that the HIT SafetyCenter leverages its unique position to work with key administrative and policy stakeholders, healthcare organizations (HCOs), and HIT vendors to achieve four goals: (1) facilitate creation of a nationwide 'post-marketing' surveillance system to monitor HIT related safety events; (2) develop methods and governance structures to support investigation of major HIT related safety events; (3) create the infrastructure and methods needed to carry out random assessments of HIT related safety in complex HCOs; and (4) advocate for HIT safety with government and private entities. The convening ability of a federally supported HIT SafetyCenter could be critically important to our transformation to a safe and effective HIT enabled healthcare system. PMID:25332353

Foodborne illness has a significant impact on public health and consumer confidence in the U.S. The Southwest Border Food Safety and Defense Center was established to provide educational programs, trainings, and workshops to address the health and well-being of consumers as it relates to food safety and food protection. A partnership between New…

This index to NuclearSafety, a bimonthly technical progress review, covers articles published in NuclearSafety, Volume II, No. 1 (January-February 1970), through Volume 20, No. 6 (November-December 1979). It is divided into three sections: a chronological list of articles (including abstracts) followed by a permuted-title (KWIC) index and an author index. NuclearSafety, a bimonthly technical progress review prepared by the NuclearSafety Information Center (NSIC), covers all safety aspects of nuclear power reactors and associated facilities. Over 600 technical articles published in NuclearSafety in the last ten years are listed in this index.

This index to NuclearSafety covers articles published in NuclearSafety, Volume 18, Number 1 (January-February 1977) through Volume 22, Number 6 (November-December 1981). The index is divided into three section: a chronological list of articles (including abstracts), a permuted-title (KWIC) index, and an author index. NuclearSafety, a bimonthly technical progress review prepared by the NuclearSafety Information Center, covers all safety aspects of nuclear power reactors and associated facilities. Over 300 technical articles published in NuclearSafety in the last 5 years are listed in this index.

The Geel Electron Linear Accelerator (GELINA) was used to measure neutron total and capture cross sections of {sup 182,183,184,186}W and {sup 63,65}Cu in the energy range from 100 eV to {approx}200 keV using the time-of-flight method. GELINA is the only high-power white neutron source with excellent timing resolution and ideally suited for these experiments. Concerns about the use of existing cross-section data in nuclear criticality calculations using Monte Carlo codes and benchmarks were a prime motivator for the new cross-section measurements. To support the Nuclear Criticality Safety Program, neutron cross-section measurements were initiated using GELINA at the EC-JRC-IRMM. Concerns about data deficiencies in some existing cross-section evaluations from libraries such as ENDF/B, JEFF, or JENDL for nuclear criticality calculations were the prime motivator for new cross-section measurements. Over the past years many troubles with existing nuclear data have emerged, such as problems related to proper normalization, neutron sensitivity backgrounds, poorly characterized samples, and use of improper pulse-height weighting functions. These deficiencies may occur in the resolved- and unresolved-resonance region and may lead to erroneous nuclear criticality calculations. An example is the use of the evaluated neutron cross-section data for tungsten in nuclear criticality safety calculations, which exhibit discrepancies in benchmark calculations and show the need for reliable covariance data. We measured the neutron total and capture cross sections of {sup 182,183,184,186}W and {sup 63,65}Cu in the neutron energy range from 100 eV to several hundred keV. This will help to improve the representation of the cross sections since most of the available evaluated data rely only on old measurements. Usually these measurements were done with poor experimental resolution or only over a very limited energy range, which is insufficient for the current application.

The Hanford Database is a useful information retrieval tool for a criticality safety practitioner. The database contains nuclear criticality literature screened for parameter studies. The entries, characterized with a value index, are segregated into 16 major and six minor categories. A majority of the screened entries have abstracts and a limited number are connected to the Office of Scientific and Technology Information (OSTI) database of full-size documents. Simple and complex searches of the data can be accomplished very rapidly and the end-product of the searches could be a full-size document. The paper contains a description of the database, user instructions, and a number of examples.

This report discusses the results of research to evaluate existing and/or recommended safety/reliability assurance activities among nuclear and other high technology industries for potential nuclear industry implementation. Since the Three Mile Island (TMI) accident, there has been increased interest in the use of reliability programs (RP) to assure the performance of nuclearsafety systems throughout the plant's lifetime. Recently, several Nuclear Regulatory Commission (NRC) task forces or safety issue review groups have recommended RPs for assuring the continuing safety of nuclear reactor plants. 18 references.

This report discusses the necessity for in-depth safety program planning for space nuclear reactor systems. The objectives of the safety program and a proposed task structure is presented for meeting those objectives. A proposed working relationship between the design and independent safety groups is suggested. Examples of safety-related design philosophies are given.

This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

A survey is presented of some current trends in nuclearsafety analysis that may be relevant to space nuclear power systems. This includes: lessons learned from operating power reactor safety and licensing; approaches to the safety design of advanced and novel reactors and facilities; the roles of risk assessment, extremely unlikely accidents, safety goals/targets; and risk-benefit analysis and communication.

Safety is the prime design requirement for nuclear thermal propulsion (NTP). It must be built in at the initiation of the design process. An understanding of safety concerns is fundamental to the development of nuclear rockets for manned missions to Mars and many other applications that will be enabled or greatly enhanced by the use of nuclear propulsion. To provide an understanding of the basic issues, a tutorial has been prepared. This tutorial covers a range of topics including safety requirements and approaches to meet these requirements, risk and safety analysis methodology, NERVA reliability and safety approach, and life cycle risk assessments.

Safety is the prime design requirement for nuclear thermal propulsion (NTP). It must be built in at the initiation of the design process. An understanding of safety concerns is fundamental to the development of nuclear rockets for manned missions to Mars and many other applications that will be enabled or greatly enhanced by the use of nuclear propulsion. To provide an understanding of the basic issues, a tutorial has been prepared. This tutorial covers a range of topics including safety requirements and approaches to meet these requirements, risk and safety analysis methodology, NERVA reliability and safety approach, and life cycle risk assessments.

The IAEA Safety Standards Series include, in a hierarchical manner, the categories of Safety Fundamentals, Safety Requirements and Safety Guides, which define the elements necessary to ensure the safety of nuclear installations. In the same way as nuclear technology and scientific knowledge advance continuously, also safety requirements may change with these advances. Therefore, in the framework of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) one important aspect among others refers to user requirements on the safety of innovative nuclear installations, which may come into operation within the next fifty years. In this respect, the major objectives of the INPRO sub-task 'User Requirements and Nuclear Energy Development Criteria in the Area of Safety' have been: a. to overview existing national and international requirements in the safety area, b. to define high level user requirements in the area of safety of innovative nuclear technologies, c. to compile and to analyze existing innovative reactor and fuel cycle technology enhancement concepts and approaches intended to achieve a high degree of safety, and d. to identify the general areas of safety R and D needs for the establishment of these technologies. During the discussions it became evident that the application of the defence in depth strategy will continue to be the overriding approach for achieving the general safety objective in nuclear power plants and fuel cycle facilities, where the emphasis will be shifted from mitigation of accident consequences more towards prevention of accidents. In this context, four high level user requirements have been formulated for the safety of innovative nuclear reactors and fuel cycles. On this basis safety strategies for innovative reactor designs are highlighted in each of the five levels of defence in depth and specific requirements are discussed for the individual components of the fuel cycle. (authors)

Results from an extensive aging program on polymeric materials stripped from unused nuclear reactor safety cables are described. Mechanical damage was monitored after room temperature aging in a Co-60 gamma radiation source at various humidities and radiation dose rates ranging from 1.2 Mrad/h to 2 krad/h. For chloroprene, chlorosulfonated polyethylene, and silicone materials, the mechanical degradation was found to depend only on the total integrated radiation dose, implying that radiation dose rate effects are small. On the other hand, strong evidence for radiation dose rate effects were found for an ethylene propylene rubber material and a cross-linked polyolefin material. Humidity effects were determined to be insignificant for all the materials studied.

With the completion of the Space Shuttle Program, the Kennedy Space Center (KSC) safety function will be required to evolve beyond the single launch vehicle launch site focus that has held prominence for almost fifty years. This paper will discuss how that evolution is taking place. Specifically, we will discuss the future of safety as it relates to a site that will have multiple, very disparate, functions. These functions will include new business; KSC facilities not under the control of NASA; traditional payload and launch vehicle processing; and, operations conducted by NASA personnel, NASA contractors or a combination of both. A key element in this process is the adaptation of the current KSC set of safety requirements into a multi-faceted set that can address each of the functions above, while maintaining our world class safety environment. One of the biggest challenges that will be addressed is how to protect our personnel and property without dictating how other Non-NASA organizations protect their own employees and property. The past history of KSC Safety will be described and how the lessons learned from previous programs will be applied to the future. The lessons learned from this process will also be discussed as information for other locations that may undergo such a transformation.

The Accurate fission data for nuclearsafety (AlFONS) project aims at high precision measurements of fission yields, using the renewed IGISOL mass separator facility in combination with a new high current light ion cyclotron at the University of Jyväskylä. The 30 MeV proton beam will be used to create fast and thermal neutron spectra for the study of neutron induced fission yields. Thanks to a series of mass separating elements, culminating with the JYFLTRAP Penning trap, it is possible to achieve a mass resolving power in the order of a few hundred thousands. In this paper we present the experimental setup and the design of a neutron converter target for IGISOL. The goal is to have a flexible design. For studies of exotic nuclei far from stability a high neutron flux (1012 neutrons/s) at energies 1 - 30 MeV is desired while for reactor applications neutron spectra that resembles those of thermal and fast nuclear reactors are preferred. It is also desirable to be able to produce (semi-)monoenergetic neutrons for benchmarking and to study the energy dependence of fission yields. The scientific program is extensive and is planed to start in 2013 with a measurement of isomeric yield ratios of proton induced fission in uranium. This will be followed by studies of independent yields of thermal and fast neutron induced fission of various actinides.

The U.S. Department of Energy, through its agents the Next Generation Nuclear Plant Project and the Nuclear Hydrogen Initiative, is working on developing the technologies to enable the large scale production of hydrogen using nuclear power. A very important consideration in the design of a co-located and connected nuclear plant/hydrogen plant facility is safety. This study provides an overview of the safety issues associated with a combined plant and discusses approaches for categorizing, quantifying, and addressing the safety risks.

One of the challenges that today's new workforce of nuclear criticality safety engineers face is the opportunity to provide assessment of nuclear systems and establish safety guidelines without having received significant experience or hands-on training prior to graduation. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and/or the International Reactor Physics Experiment Evaluation Project (IRPhEP) provides students and young professionals the opportunity to gain experience and enhance critical engineering skills.

The Nuclear Data Centers Network is a world-wide cooperation of nuclear data centers under the auspices of the International Atomic Energy Agency. The Network organizes the task of collecting, compiling, standardizing, storing, assessing, and distributing the nuclear data on an international scale. Information available at the Centers includes bibliographic, experimental, and evaluated databases for nuclear reaction data and for nuclear structure and radioactive decay data. The objective of the Network is to provide the information to users in a convenient, readily-available form. To this end, online data services have been established at three of the centers: the National Nuclear Data Center (NNDC), the Nuclear Data Section of the International Atomic Energy Agency (NDS), and the OECD Nuclear Energy Agency Data Bank (NEADB). Some information is also available at the NNDC and NEADB World Wide Web sites.

As part of the privatization process of the U.S. uranium enrichment plants, the Paducah Gaseous Diffusion Plant (PGDP) and the Portsmouth Gaseous Diffusion Plant (PORTS) have been in transition from U.S. Department of Energy (DOE) regulatory oversight to U.S. Nuclear Regulatory Commission (NRC) oversight since July 1993. One of the focus areas of this transition has been training and qualification of plant personnel who perform tasks important to nuclearsafety, such as nuclear criticality safety (NCS) engineers.

It took Westar Energy eight years to upgrade the Lawrence Energy Center to burn Powder River Basin coal. Its zero lost-time accident record during the eight-year, million-man-hour project is a testament to Westar's commitment to workplace safety. The plant won the Powder River Basin Coal Users' Group plant of the year award for 2006. The article describes all the changes implemented at the plant, including replacing and upgrading controls for the belt conveyor, replacing the coal crushers, minimising dust and modifying coal bunkers, to cope with the increased volatility of Powder River Basin coal. Modifications were made to minimise slagging and fouling of boilers. 10 photos.

The Generation IV International Forum (GIF) Risk and Safety Working Group (RSWG) was created to develop an effective approach for the safety of Generation IV advanced nuclear energy systems. Early work of the RSWG focused on defining a safety philosophy founded on lessons learned from current and prior generations of nuclear technologies, and on identifying technology characteristics that may help achieve Generation IV safety goals. More recent RSWG work has focused on the definition of an integrated safety assessment methodology for evaluating the safety of Generation IV systems. The methodology, tentatively called ISAM, is an integrated “toolkit” consisting of analytical techniques that are available and matched to appropriate stages of Generation IV system concept development. The integrated methodology is intended to yield safety-related insights that help actively drive the evolving design throughout the technology development cycle, potentially resulting in enhanced safety, reduced costs, and shortened development time.

Review of NMP-NCS-930087, {open_quotes}Nuclear Criticality Safety Evaluation 93-04 Enriched Uranium Receipt (U), July 30, 1993, {close_quotes} was requested of SRTC (Savannah River Technology Center) Applied Physics Group. The NCSE is a criticality assessment to determine the mass limit for Engineered Low Level Trench (ELLT) waste uranium burial. The intent is to bury uranium in pits that would be separated by a specified amount of undisturbed soil. The scope of the technical review, documented in this report, consisted of (1) an independent check of the methods and models employed, (2) independent HRXN/KENO-V.a calculations of alternate configurations, (3) application of ANSI/ANS 8.1, and (4) verification of WSRC Nuclear Criticality Safety Manual procedures. The NCSE under review concludes that a 500 gram limit per burial position is acceptable to ensure the burial site remains in a critically safe configuration for all normal and single credible abnormal conditions. This reviewer agrees with that conclusion.

Current space nuclear power reactor safety issues are discussed with respect to the unique characteristics of these reactors. An approach to achieving adequate safety and a perception of safety is outlined. This approach calls for a carefully conceived safety program which makes uses of lessons learned from previous terrestrial power reactor development programs. This approach includes use of risk analyses, passive safety design features, and analyses/experiments to understand and control off-design conditions. The point is made that some recent accidents concerning terrestrial power reactors do not imply that space power reactors cannot be operated safety.

The nuclearsafety integration and operational aspects of transporting nuclear payloads to and from an earth orbiting space base by space shuttle are discussed. The representative payloads considered were: (1) zirconium hydride-Brayton power module, (2) isotope-Brayton power module, and (3) small isotope power systems or heat sources. Areas of investigation also include nuclearsafety related integration and packaging as well as operational requirements for the shuttle and payload systems for all phases of the mission.

Little guidance has been provided by the DOE for developing appropriate Operational Safety Requirements (OSR) for non-nuclear facility safety documents. For a period of time, Chapter 2 of DOE/AL Supplemental Order 5481.lB provided format guidance for non-reactor nuclear facility OSRs when this supplemental order applied to both nuclear and non-nuclear facilities. Thus, DOE Albuquerque Operations Office personnel still want to see non-nuclear facility OSRs in accordance with the supplemental order (i.e., in terms of Safety Limits, Limiting Conditions for Operation, and Administrative Controls). Furthermore, they want to see a clear correlation between the OSRs and the results of a facility safety analysis. This paper demonstrates how OSRs can be rather simply derived from the results of a risk assessment performed using the ``binning`` methodology of SAND95-0320.

This report documents the US Nuclear Regulatory Commission (NRC) staff review and safety evaluation of the Louisiana Energy Services, L.P. (LES, the applicant) application for a license to possess and use byproduct, source, and special nuclear material and to enrich natural uranium to a maximum of 5 percent U-235 by the gas centrifuge process. The plant, to be known as the Claiborne Enrichment Center (CEC), would be constructed near the town of Homer in Claiborne Parish, Louisiana. At full production in a given year, the plant will receive approximately 4,700 tonnes of feed UF{sub 6} and produce 870 tonnes of low-enriched UF{sub 6}, and 3,830 tonnes of depleted UF{sub 6} tails. Facility construction, operation, and decommissioning are expected to last 5, 30, and 7 years, respectively. The objective of the review is to evaluate the potential adverse impacts of operation of the facility on worker and public health and safety under both normal operating and accident conditions. The review also considers the management organization, administrative programs, and financial qualifications provided to assure safe design and operation of the facility. The NRC staff concludes that the applicant`s descriptions, specifications, and analyses provide an adequate basis for safety review of facility operations and that construction and operation of the facility does not pose an undue risk to public health and safety.

This document is a compilation and source list of nuclearsafety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclearsafety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

The NuclearSafety Working Group of the Pacific Nuclear Council promotes nuclearsafety cooperation among its members. Status of safety research, emergency planning, development of lists of technical experts, severe accident prevention and mitigation have been the topics of discussion in the NSWG. This paper reviews and compares the severe accident prevention and mitigation program activities in some of the areas of the Pacific Basin region based on papers presented at a special session organized by the NSWG at an ANS Topical Meeting as well as papers from other sources.

The NuclearSafety Working Group of the Pacific Nuclear Council promotes nuclearsafety cooperation among its members. Status of safety research, emergency planning, development of lists of technical experts, severe accident prevention and mitigation have been the topics of discussion in the NSWG. This paper reviews and compares the severe accident prevention and mitigation program activities in some of the areas of the Pacific Basin region based on papers presented at a special session organized by the NSWG at an ANS Topical Meeting as well as papers from other sources.

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclearsafety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclearsafety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

The Chart of Nuclides is a software product that allows users to search and plot nuclear structure and nuclear decay data interactively. The Chart of Nuclides was developed by the National Nuclear Data Center (NNDC). It provides an interface between web users and several databases containing nuclear structure, nuclear decay and some neutron-induced nuclear reaction information. Using the Chart of Nuclides, it is possible to search for nuclear level properties (energy, half-life, spin-parity), gamma-ray information (energy, intensity, multipolarity, coincidences),radiation information following nuclear decay (energy, intensity, dose), and neutron-induced reaction data from the BNL-325 book (thermal cross section and resonance integral). The information provided by the Chart of Nuclides can be seen in tables, level schemes and an interactive chart of nuclides. (From the Chart of Nuclides Description at http://www.nndc.bnl.gov/chart/help/index.jsp?product=chart)

The future of nuclear energy could depend upon the international infrastructure established to ensure the creation of a strong and uniform safety culture. Deliberations during the 1990s, leading to the recently promulgated International NuclearSafety Convention, held out the prospect of both bolstering nuclearsafety and gaining public recognition of the need to address transboundary safety concerns head-on. Unfortunately, the Convention that emerged from the deliberations constitutes little more than another form of technical assistance. The basis for an alternative, and more substantial, Convention is presented--one that would be based on the establishment and evaluation of performance standards, the creation of a series of political firebreaks, and the encouragement of nuclear power plant designs that minimize the catastrophic offsite consequences of accidents.

Seven Nuclear Regulatory Commission (NRC) engineers have called on the commission to force the owners of US nuclear reactors to repair a design flaw that could affect the safe operation of emergency core cooling systems.

In fission yeast, microtubules push against the cell edge, thereby positioning the nucleus in the cell center. Kinesin-8 motors regulate microtubule catastrophe; however, their role in nuclear positioning is not known. Here we develop a physical model that describes how kinesin-8 motors affect nuclearcentering by promoting a microtubule catastrophe. Our model predicts the improved centering of the nucleus in the presence of motors, which we confirmed experimentally in living cells. The model also predicts a characteristic time for the recentering of a displaced nucleus, which is supported by our experiments where we displaced the nucleus using optical tweezers.

Fifteen nuclear reactors of the type that exploded at Chernobyl in April 1986 are still operating in Russia, Ukraine, and Lithuania. The West, concerned about safety of operations, wants these reactors shut down, but the host nations refuse. The electricity these reactors supply is nuch too important for their economies, so the argument goes. The report defines policy options and procedures to implement those options for the acceptable resolution of the nuclear power safety issues facing the former Soviet Union.

The Panel on Human Factors Research Needs in Nuclear Regulatory Research was formed by the National Research Council in response to a request from the Nuclear Regulatory Commission (NRC). The NRC asked the research council to conduct an 18-month study of human factors research needs for the safe operation of nuclear power plants. This report…

These presentation visuals introduce the National Center for Nuclear Security. Its chartered mission is to enhance the Nation’s verification and detection capabilities in support of nuclear arms control and nonproliferation through R&D activities at the NNSS. It has three focus areas: Treaty Verification Technologies, Nonproliferation Technologies, and Technical Nuclear Forensics. The objectives of nuclear forensics are to reduce uncertainty in the nuclear forensics process & improve the scientific defensibility of nuclear forensics conclusions when applied to nearsurface nuclear detonations. Research is in four key areas: Nuclear Physics, Debris collection and analysis, Prompt diagnostics, and Radiochemistry.

The NASA Engineering and SafetyCenter (NESC) is an independent organization, which was charted in the wake of the Space Shuttle Columbia accident to serve as an Agency-wide technical resource focused on engineering excellence. The objective of the NESC is to improve safety by performing in-depth independent engineering assessments, testing, and analysis to uncover technical vulnerabilities and to determine appropriate preventative and corrective actions for problems, trends or issues within NASA's programs, projects and institutions. Critical to the NESC are teams of experts in a number of core disciplines including nondestructive evaluation (NDE). These teams, designated Super Problem Resolution Teams (SPRTs), draw upon the best engineering expertise from across the Agency and include partnerships with other government agencies, national laboratories, universities and industry. The NESC NDE SPRT provides a ready resource of NDE technical expertise to support NESC Independent Technical Assessments and Investigations. The purpose of this session will be to provide an overview of the NESC and the NDE SPRT along with a few examples of NDE related problems that the team has addressed for NASA Programs. It is hoped that this session will be of interest to the general NDE community and will foster contacts with additional NDE experts that might provide future support to the NASA NESC NDE SPRT.

Insights developed within the U.S. nuclear weapon system safety community may benefit system safety design, assessment, and management activities in other high consequence domains. The approach of assured nuclear weapon safety has been developed that uses the NuclearSafety Design Principles (NSDPs) of incompatibility, isolation, and inoperability to design safety features, organized into subsystems such that each subsystem contributes to safe system responses in independent and predictable ways given a wide range of environmental contexts. The central aim of the approach is to provide a robust technical basis for asserting that a system can meet quantitative safety requirements in the widest context of possible adverse or accident environments, while using the most concise arrangement of safety design features and the fewest number of specific adverse or accident environment assumptions. Rigor in understanding and applying the concept of independence is crucial for the success of the approach. This paper provides a basic description of the assured nuclear weapon safety approach, in a manner that illustrates potential application to other domains. There is also a strong emphasis on describing the process for developing a defensible technical basis for the independence assertions between integrated safety subsystems.

A turning point of the world nuclear industry with respect to safety occurred due to the accident at Chernobyl, in 1986. A side from the tragic personal losses and the enormous financial damage, the Chernobyl accident has literally demonstrated that ''a nuclear accident anywhere is an accident everywhere''. The impact was felt immediately by the nuclear industry, with plant cancellations (e.g. Austria), elimination of national programs (e.g. Italy) and general construction delays. However, the reaction of the nuclear industry was equally immediate, which led to the proposal and establishment of a Global NuclearSafety Regime. This regime is composed of biding international safety conventions, globally accepted safety standard, and a voluntary peer review system. In a previous work, the author has presented in detail the components of this Regime, and briefly discussed its impact in the Brazilian nuclear power organizations, including the Regulatory Body. This work, on the opposite, briefly reviews the Global NuclearSafety Regime, and concentrates in detail in the discussion of its impact in Brazil, showing how it has produced some changes, and where the peer pressure regime has failed to produce real results.

The National Nuclear Data Center (NNDC) collects, evaluates, and disseminates nuclear physics data for basic research and applied nuclear technologies. In 2004, the NNDC migrated all databases into modern relational database software, installed new generation of Linux servers and developed new Java-based Web service. This nuclear database development means much faster, more flexible and more convenient service to all users in the United States. These nuclear reaction and structure database developments as well as related Web services are briefly described.

We discuss nuclear data resources of the National Nuclear Data Center (NNDC) of relevance to nuclear astrophysics applications. These resources include databases, tools and powerful web service at www.nndc.bnl.gov. Our objective is to provide an overview of nuclear databases, related products and demonstrate nuclear astrophysics potential of the ENDF/B-VII beta2 library. A detailed discussion on the Maxwellian neutron capture cross sections obtained from the ENDF/B-VII beta2 library is presented.

A year after the accident at Three Mile Island (TMI), the signals transmitted to the public are still confused. Industry says that nuclear power is safe and that the aftermath of TMI ushers in a new era of safety. Antinuclear activists say TMI sounded nuclear power's death knell. (Author/RE)

A bibliographic criticality safety database of over 13,000 records is available on the Internet as part of the U.S. Department of Energy's (DOE) Nuclear Criticality Safety Program (NCSP) website. This database is easy to access via the Internet and gets substantial daily usage. This database and other criticality safety resources are available at ncsp.llnl.gov. The web database has evolved from more than thirty years of effort at Lawrence Livermore National Laboratory (LLNL), beginning with compilations of critical experiment reports and American Nuclear Society Transactions.

EDF has conducted, after the Fukushima event, complementary safety assessments of its nuclear facilities. The aim of this in-depth review was to assess the resilience of each plant to extreme external hazards, situations that could lead to severe accident conditions. These analyses demonstrate a good level of safety for all of EDF's nuclear facilities. Supplementary measures post-Fukushima have been put forward to the ASN with the aim of continuing to improve the level of safety at the plants. Once the ASN position is issued, EDF will develop an action plan over several years, covering both supplementary studies and modifications that have been identified.

Activities and infrastructure that support nuclear weapons are facing significant challenges. Despite an admirable record and firm commitment to make safety a primary criterion in weapons design, production, handling, and deployment - there is growing apprehension about terrorist acquiring weapons or nuclear material. At the NES Workshop in May 2012, Scott Sagan, who is a proponent of the normal accident cycle, presented. Whether a proponent of the normal accident cycle or High Reliability Organizations - we have to be diligent about our safety record. Constant vigilance is necessary to maintain our admirable safety record and commitment to Nuclear Explosive Safety.

The results of a study to identify potential hazards arising from nuclear reactor power systems for use on the lunar and Martian surfaces, related safety issues, and resolutions of such issues by system design changes, operating procedures, and other means are presented. All safety aspects of nuclear reactor power systems from prelaunch ground handling to eventual disposal were examined consistent with the level of detail for SP-100 reactor design at the 1988 System Design Review and for launch vehicle and space transport vehicle designs and mission descriptions as defined in the 90-day Space Exploration Initiative (SEI) study. Information from previous aerospace nuclearsafety studies was used where appropriate. Safety requirements for the SP-100 space nuclear reactor system were compiled. Mission profiles were defined with emphasis on activities after low earth orbit insertion. Accident scenarios were then qualitatively defined for each mission phase. Safety issues were identified for all mission phases with the aid of simplified event trees. Safety issue resolution approaches of the SP-100 program were compiled. Resolution approaches for those safety issues not covered by the SP-100 program were identified. Additionally, the resolution approaches of the SP-100 program were examined in light of the moon and Mars missions.

This document contains summaries of most of the papers presented at the 1995 Nuclear Criticality Technology Safety Project (NCTSP) meeting, which was held May 16 and 17 at San Diego, Ca. The meeting was broken up into seven sessions, which covered the following topics: (1) Criticality Safety of Project Sapphire; (2) Relevant Experiments For Criticality Safety; (3) Interactions with the Former Soviet Union; (4) Misapplications and Limitations of Monte Carlo Methods Directed Toward Criticality Safety Analyses; (5) Monte Carlo Vulnerabilities of Execution and Interpretation; (6) Monte Carlo Vulnerabilities of Representation; and (7) Benchmark Comparisons.

... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF HEALTH AND HUMAN SERVICES Centers for Disease Control and Prevention National Institute for Occupational Safety and Health, (NIOSH), World Trade Center Health Program Science/Technical Advisory Committee...

The recent nuclear renaissance has motivated many countries, especially developing nations, to plan and build nuclear power reactors. However, domestic low enriched uranium demands may trigger nations to construct indigenous enrichment facilities, which could be redirected to fabricate high enriched uranium for nuclear weapons. The potential advantages of establishing multinational uranium enrichment sites are numerous including increased low enrichment uranium access with decreased nuclear proliferation risks. While multinational nuclear initiatives have been discussed, Russia is the first nation to actualize this concept with their Angarsk International Uranium Enrichment Center (IUEC). This paper provides an overview of the historical and modern context of the multinational nuclear fuel cycle as well as the evolution of Russia's IUEC, which exemplifies how international fuel cycle cooperation is an alternative to domestic facilities.

The purpose of this document is to define safety criteria which must be met to implement US safety policy for space fission reactors. These criteria provide the bases for decisions on the acceptability of specific mission and reactor design proposals. (JDH)

This report discusses aging effects on safety-related power transformers in nuclear power plants. It also evaluates maintenance, testing, and monitoring practices with respect to their effectiveness in detecting and mitigating the effects of aging. The study follows the US Nuclear Regulatory Commission`s (NRC`s) Nuclear Plant-Aging Research approach. It investigates the materials used in transformer construction, identifies stressors and aging mechanisms, presents operating and testing experience with aging effects, analyzes transformer failure events reported in various databases, and evaluates maintenance practices. Databases maintained by the nuclear industry were analyzed to evaluate the effects of aging on the operation of nuclear power plants.

The present special issue contains an overview of the research in the field of Integrated Deterministic and Probabilistic Safety Assessment (IDPSA) of Nuclear Power Plants (NPPs). Traditionally, safety regulation for NPPs design and operation has been based on Deterministic Safety Assessment (DSA) methods to verify criteria that assure plant safety in a number of postulated Design Basis Accident (DBA) scenarios. Referring to such criteria, it is also possible to identify those plant Structures, Systems, and Components (SSCs) and activities that are most important for safety within those postulated scenarios. Then, the design, operation, and maintenance of these “safety-related” SSCs andmore » activities are controlled through regulatory requirements and supported by Probabilistic Safety Assessment (PSA).« less

The present special issue contains an overview of the research in the field of Integrated Deterministic and Probabilistic Safety Assessment (IDPSA) of Nuclear Power Plants (NPPs). Traditionally, safety regulation for NPPs design and operation has been based on Deterministic Safety Assessment (DSA) methods to verify criteria that assure plant safety in a number of postulated Design Basis Accident (DBA) scenarios. Referring to such criteria, it is also possible to identify those plant Structures, Systems, and Components (SSCs) and activities that are most important for safety within those postulated scenarios. Then, the design, operation, and maintenance of these “safety-related” SSCs and activities are controlled through regulatory requirements and supported by Probabilistic Safety Assessment (PSA).

Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems. PMID:18049233

This document summarizes and describes how to access the on-line services available from the National Nuclear Data Center (NNDC) located at Brookhaven National Laboratory. The services are available free of cost to US Department of Energy, its contractors and others who support the NNDC or supply data to the NNDC. Four of the center's data bases are now accessible to non-NNDC scientists via remote connection to the center's VAX 11/780. To use this service, you must have a terminal with access by either a telephone line or the PHYSNET network. A VT100 terminal or a terminal with VT-100 emulation is recommended but not required.

The Lawrence Livermore National Laboratory has prepared a Nuclear Criticality Safety Bibliographic Database that is now available via the Internet. This database is a component of the U.S. DOE Nuclear Criticality Safety Program (NCSP) Web site. This WWW resource was developed as part of the DOE response to the DNFSB Recommendation 97-2, which reflected the need to make criticality safety information available to a wide audience. To the extent possible, the hyperlinks on the Web pages direct the user to original source of the reference material in order to ensure accuracy and access to the latest versions. A master index is in place for simple navigation through the site. A search capability is available to assist in locating the on-line reference materials. Among the features included are: A user-friendly site map for ease of use; A personnel registry; Links to all major laboratories and organizations involved in the many aspects of criticality safety; General help for new criticality safety practitioners, including basic technical references and training modules; A discussion of computational methods; An interactive question and answer forum for the criticality safety community; and Collections of bibliographic references mdvahdation experiments. This paper will focus on the bibliographic database. This database evolved from earlier work done by the DOE's Nuclear Criticality Information System (NCIS) maintained at LLNL during the 1980s. The bibliographic database at the time of the termination of NCIS were composed principally of three parts: (1) A critical experiment bibliography of 1067 citations (reported in UCRL-52769); (2) A compilation of criticality safety papers from Volumes 1 through 41 of the Transactions of the American Nuclear Society (reported in UCRL-53369); and (3) A general criticality bibliography of several thousand citations (unpublished). When the NCIS project was terminated the database was nearly lost but, fortunately, several years later

Faced with increasing global energy demands, many developing countries are considering building their first nuclear power plant. As a country embarks upon or expands its nuclear power program, it should consider how it will address the 19 issues laid out in the International Atomic Energy Agency (IAEA) document Milestones in Development of a National Infrastructure for Nuclear Power. One of those issues specifically addresses the international nonproliferation treaties and commitments and the implementation of safeguards to prevent diversion of nuclear material from peaceful purposes to nuclear weapons. Given the many legislative, economic, financial, environmental, operational, and other considerations preoccupying their planners, it is often difficult for countries to focus on developing the core strengths needed for effective safeguards implementation. Typically, these countries either have no nuclear experience or it is limited to the operation of research reactors used for radioisotope development and scientific research. As a result, their capacity to apply safeguards and manage fuel operations for a nuclear power program is limited. This paper argues that to address the safeguards issue effectively, a holistic approach must be taken to integrate safeguards with the other IAEA issues including safety and security - sometimes referred to as the '3S' concept. Taking a holistic approach means that a country must consider safeguards within the context of its entire nuclear power program, including operations best practices, safety, and security as well as integration with its larger nonproliferation commitments. The Department of Energy/National Nuclear Security Administration's International Nuclear Safeguards and Engagement Program (INSEP) has been involved in bilateral technical cooperation programs for over 20 years to promote nonproliferation and the peaceful uses of nuclear energy. INSEP is currently spearheading efforts to promote the development of

This document provides Department of Energy (DOE) contractors, subcontractors and suppliers with guidance in the effective use of DOE`s Price-Anderson nuclearsafety Noncompliance Tracking System (NTS). Prompt contractor identification, reporting to DOE, and correction of nuclearsafety noncompliances provides DOE with a basis to exercise enforcement discretion to mitigate civil penalties, and suspend the issuance of Notices of Violation for certain violations. Use of this reporting methodology is elective by contractors; however, this methodology is intended to reflect DOE`s philosophy on effective identification and reporting of nuclearsafety noncompliances. To the extent that these expectations are met for particular noncompliances, DOE intends to appropriately exercise its enforcement discretion in considering whether, and to what extent, to undertake enforcement action.

This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed here are ongoing. Results and conclusions described may change as the work continues.

For the past two decades, Spain has been consolidating a nuclear program that in the last 3 years has provided between 35 and 40% of the electricity consumed in that country. This program includes nine operating reactor units, eight of them based on US technology and one from Germany, a total of 7,356 MW(electric). There is also a 480-MW(electric) French gas-cooled reactor whose operation recently ceased and which will be decommissioned in the coming years. Spanish industry has participated significantly in this program, and material produced locally has reached 85% of the total. Once the construction program has been completed and operation is proceeding normally, the capacity factor will be {approximately} 80%. It will be very important to complete the nuclear program with the establishment of conditions for safe management and disposal of the nuclear waste generated during the years in which these reactors are in operation and for subsequent decommissioning. To establish the guidelines for the disposal of nuclear waste, the Spanish government approved in october 1987, with a revision in January 1989, the General Plan of Radioactive Wastes proposed by the Ministry of Industry and Energy and prepared by the national company for radioactive waste management, ENRESA.

In 1975, Congress created the Nuclear Regulatory Commission (NRC). Its primary responsibility was to be the regulation of the nuclear power industry in order to maintain public health and safety. On March 28, 1979, in the worst commercial nuclear accident in US history, the plant at Three Mile Island began to leak radioactive material. How was Three Mile Island possible. Where was the NRC. This analysis by the Union of Concerned Scientists (UCS) of the NRC's first decade, points specifically to the factors that contributed to the accident at Three Mile Island. The NRC, created as a watchdog of the nuclear power industry, suffers from problems of mindset, says the UCS. The commission's problems are political, not technical; it repeatedly ranks special interests above the interest of public safety. This book critiques the NRC's performance in four specific areas. It charges that the agency has avoided tackling the most pervasive safety issues; has limited public participation in decision making and power plant licensing; has failed to enforce safety standards or conduct adequate regulation investigations; and, finally, has maintained a fraternal relationship with the industry it was created to regulate, serving as its advocate rather than it adversary. The final chapter offers recommendations for agency improvement that must be met if the NRC is to fulfill its responsibility for safety first.

PBMR is a high-temperature helium-cooled graphite-moderated continuous-fuelled pebble bed reactor. The power conversion unit is directly coupled to the reactor and the power turbines are driven through a direct closed-circuit helium cycle. An overview is presented on the nuclear engineering analyses used for the design and safety assessment for the PBMR. Topics addressed are the PBMR design, safety and licensing requirements, nuclear engineering analysis results, software verification and validation, and advances in software development. (authors)

The Nuclear Criticality Safety Organization (NCSO) is committed to developing and maintaining a staff of qualified personnel to meet the current and anticipated needs in Nuclear Criticality Safety (NCS) at the Oak Ridge Y-12 Plant. This document provides a listing of the roles and responsibilities of NCSO personnel with respect to training and details of the Training Management System (TMS) programs, Mentoring Checklists and Checksheets, as well as other documentation utilized to implement the program. This Training Implementation document is applicable to all technical and managerial NCSO personnel, including temporary personnel, sub-contractors and/or LMES employees on loan to the NCSO, who are in a qualification program.

The Los Alamos Neutron Science Center (LANSCE) facilities for Nuclear Science consist of a high-energy "white" neutron source (Target 4) with 6 flight paths, three low-energy nuclear science flight paths at the Lujan Center, and a proton reaction area. The neutron beams produced at the Target 4 complement those produced at the Lujan Center because they are of much higher energy and have shorter pulse widths. The neutron sources are driven by the 800-MeV proton beam of the LANSCE linear accelerator. With these facilities, LANSCE is able to deliver neutrons with energies ranging from a milli-electron volt to several hundreds of MeV, as well as proton beams with a wide range of energy, time and intensity characteristics. The facilities, instruments and research programs are described briefly.

Center Director Roy Bridges opens the second Super Safety and Health Day at Kennedy Space Center, an entire day when most normal work activities are suspended to allow personnel to attend safety- and health-related activities. The theme, 'Safety and Health Go Hand in Hand,' emphasized KSC's commitment to place the safety and health of the public, astronauts, employees and space- related resources first and foremost. Events included a keynote address, a panel session about related issues, vendor exhibits, and safety training in work groups. The keynote address and panel session were also broadcast internally over NASA television.

To develop a strategy for incorporating and demonstrating safety, it is necessary to enumerate the unique aspects of space power reactor systems from a safety standpoint. These features must be differentiated from terrestrial nuclear power plants so that our experience can be applied properly. Some ideas can then be developed on how safe designs can be achieved so that they are safe and perceived to be safe by the public. These ideas include operating only after achieving a stable orbit, developing an inherently safe design, ''designing'' in safety from the start and managing the system development (design) so that it is perceived safe. These and other ideas are explored further in this paper.

The findings of the Haynes Task Force Committee (NUREG-0992) are used as the basis for defining safety/security assessment team activities at commercial nuclear power plants in NRC Region V. A safety/security interface assessment outline and the approach used for making the assessments are presented along with the composition of team members. As a result of observing simulated plant emergency conditions during scheduled emergency preparedness exercises, examining security and operational response procedures, and interviewing plant personnel, the team has identified instances where safety/security conflicts can occur. 2 refs.

As part of its duties in connection with space missions involving nuclear power sources, the Office of NuclearSafety (ONS) of the Office of Assistant Secretary for Environmental Protection, Safety, and Emergency Preparedness has been assigned the task of reviewing the Overall Safety Manual (OSM) (memo from B.J. Rock to J.R. Maher, December 1, 1982). The OSM, dated July 1981 and in four volumes, was prepared by NUS Corporation, Rockville, Maryland, for the US Department of Energy. The OSM provides many of the technical models and much of the data which are used by (1) space launch contractors in safety analysis reports and (2) the broader Interagency NuclearSafety Review Panel (INSRP) safety evaluation reports. If fhs interaction between the OSM, contractors, and INSRP is to work effectively, the OSM must be accurate, comprehensive, understandable, and usable.

The Nuclear Regulatory Commission (NRC) is amending its regulations regarding hospitals licensed to use radioactive byproduct material for human applications. Currently, such a license requires that the hospital have a Medical Isotopes Committee to review clinical aspects of the use of radioactive materials within the hospital. The amendment requires instead a Radiation Safety Committee with a simplified membership that will focus on the radiation safety of workers and the general public. The rule change acknowledges the Food and Drug Administration's role in regulating the safety and effectiveness of radioactive drugs with respect to the patient. The membership of the new Radiation Safety Committee will include the hospital management and the nursing staff in decisions affecting radiation safety at the hospital and will be easier for smaller hospitals to recruit. PMID:10259789

This report addresses safety guidelines for space nuclear reactor power missions and was prepared by The Johns Hopkins University Applied Physics Laboratory (JHU/APL) under a Department of Energy grant, DE-FG01-94NE32180 dated 27 September 1994. This grant was based on a proposal submitted by the JHU/APL in response to an {open_quotes}Invitation for Proposals Designed to Support Federal Agencies and Commercial Interests in Meeting Special Power and Propulsion Needs for Future Space Missions{close_quotes}. The United States has not launched a nuclear reactor since SNAP 10A in April 1965 although many Radioisotope Thermoelectric Generators (RTGs) have been launched. An RTG powered system is planned for launch as part of the Cassini mission to Saturn in 1997. Recently the Ballistic Missile Defense Office (BMDO) sponsored the Nuclear Electric Propulsion Space Test Program (NEPSTP) which was to demonstrate and evaluate the Russian-built TOPAZ II nuclear reactor as a power source in space. As of late 1993 the flight portion of this program was canceled but work to investigate the attributes of the reactor were continued but at a reduced level. While the future of space nuclear power systems is uncertain there are potential space missions which would require space nuclear power systems. The differences between space nuclear power systems and RTG devices are sufficient that safety and facility requirements warrant a review in the context of the unique features of a space nuclear reactor power system.

This report presents an overview of effectiveness of management control of safety. It reviews several modern management control theories as well as the general functions of management and relates them to safety issues at the corporate and at the process safety management (PSM) program level. Following these discussions, structured technique for assessing management of the safety function is suggested. Seven modern management control theories are summarized, including business process reengineering, the learning organization, capability maturity, total quality management, quality assurance and control, reliability centered maintenance, and industrial process safety. Each of these theories is examined for-its principal characteristics and implications for safety management. The five general management functions of planning, organizing, directing, monitoring, and integrating, which together provide control over all company operations, are discussed. Under the broad categories of Safety Culture, Leadership and Commitment, and Operating Excellence, key corporate safety elements and their subelements are examined. The three categories under which PSM program-level safety issues are described are Technology, Personnel, and Facilities.

For implementation of advance passive safety features in future nuclear power plant design, a passive safety system has been proposed and its response has been observed for Loss of Coolant Accident (LOCA) in the cold leg of a reactor coolant system. In a transient simulation the performance of proposed system is validated against existing safety injection system for a reference power plant of 325 MWe. The existing safety injection system is a huge system and consists of many active components including pumps, valves, piping and Instrumentation and Control (I&C). A good running of the active components of this system is necessary for its functionality as high head safety injection system under design basis accidents. Using reactor simulation technique, the proposed passive safety injection system and existing safety injection system are simulated and tested for their performance under large break LOCA for the same boundary conditions. Critical thermal hydraulic parameters of both the systems are presented graphically and discussed. The results obtained are approximately the same in both the cases. However, the proposed passive safety injection system is a better choice for such type of reactors due to reduction in components with improved safety.

Considerable experience has been accumulated in the Soviet Union on the design, construction and operation of nuclear-powered civilian ships: the icebreakers Lenin, Leonid Brezhnev and Sibir. The nuclear steam plants (NSP) used on these as the main energy source have been found to be highly reliable and safe, and it is desirable to use them in the future not only in icebreakers but also in transport ships for use in ice fields. The Soviet program for building and developing nuclear-powered ships has involved careful attention to safety in ships containing NSP. The experience with the design and operation of nuclear icebreakers in recent years has led to the revision of safety standards for the nuclear ships and correspondingly ship NSP and international guidelines have been developed. If one meets the requirements as set forth in these documents, one has a safe basis for future Soviet nuclear-powered ships. The primary safety provisions for NSP are presented in this paper.

This is a document required by Basic Energy Sciences as part of a mid-term review, in the third year of the five-year award period and is intended to provide a critical assessment of the Center for Materials Science of Nuclear Fuels (strategic vision, scientific plans and progress, and technical accomplishments).

The Nuclear Forensics Analysis Center (NFAC) is part of Savannah River National Laboratory (SRNL) and is one of only two USG National Laboratories accredited to perform nuclear forensic analyses to the requirements of ISO 17025. SRNL NFAC is capable of analyzing nuclear and radiological samples from bulk material to ultra-trace samples. NFAC provides analytical support to the FBI's Radiological Evidence Examination Facility (REEF), which is located within SRNL. REEF gives the FBI the capability to perform traditional forensics on material that is radiological and/or is contaminated. SRNL is engaged in research and development efforts to improve the USG technical nuclear forensics capabilities. Research includes improving predictive signatures and developing a database containing comparative samples.

Records of the Navy SafetyCenter, Norfolk, VA were reviewed to find data relevant to inadvertant operation of installed fire extinguishing systems in civilian nuclear power plants. Navy data show the incidence of collateral fire or other damage by fresh water on operating electrical equipment in submarines and in shore facilities is about the same as the civilian experience, about 30 percent. Aboard surface ships, however, the collateral damage incidence in much lower, about 15 percent. With sea water, the collateral damage incidence is at least 75 percent. It is concluded that the fire extinguisher water has to be contaminated, as by rust in sprinkler systems or deposited salt spray, for most collateral damage to occur. Reasons for inadvertant operation (or advertant operation) of firex systems at shore facilities, submarines, and surface ships resemble those for nuclear power plants. Mechanical or electrical failures lead the list, followed by mishaps during maintenance. Detector and alarm system failures are significant problems at Navy shore facilities, and significant at nuclear power plants. Fixed halon and CO2 systems in shore facilities cause no collateral damage. Lists of individual Navy incidents with water and with halon and carbon dioxide are included as appendices.

The site specific analysis of nuclear criticality training needs is very briefly described. Analysis indicated that the four major components required were analysis, surveillance, business practices or administration, and emergency preparedness. The analysis component was further divided into process analysis, accident analysis, and transportation analysis. Ten subject matter areas for the process analysis component were identified as candidates for class development. Training classes developed from the job content analysis have demonstrated that the specialized information can be successfully delivered to new entrants. 1 fig.

This paper describes the design and procurement of a Class IE DC power supply system to upgrade plant safety at the Kola Nuclear Power Plant (NPP). Kola NPP is located above the Arctic circle at Polyarnie Zorie, Murmansk, Russia. Kola NPP consists of four units. Units 1 and 2 have VVER-440/230 type reactors: Units 3 and 4 have VVER-440/213 type reactors. The VVER-440 reactor design is similar to the pressurized water reactor design used in the US. This project provided redundant, Class 1E DC station batteries and DC switchboards for Kola NPP, Units 1 and 2. The new DC power supply system was designed and procured in compliance with current nuclear design practices and requirements. Technical issues that needed to be addressed included reconciling the requirements in both US and Russian codes and satisfying the requirements of the Russian nuclear regulatory authority. Close interface with ATOMENERGOPROEKT (AEP), the Russian design organization, KOLA NPP plant personnel, and GOSATOMNADZOR (GAN), the Russian version of US Nuclear Regulatory Commission, was necessary to develop a design that would assure compliance with current Russian design requirements. Hence, this project was expected to serve as an example for plant upgrades at other similar VVER-440 nuclear plants. In addition to technical issues, the project needed to address language barriers and the logistics of shipping equipment to a remote section of the Former Soviet Union (FSU). This project was executed by Burns and Roe under the sponsorship of the US DOE as part of the International Safety Program (INSP). The INSP is a comprehensive effort, in cooperation with partners in other countries, to improve nuclearsafety worldwide. A major element within the INSP is the improvement of the safety of Soviet-designed nuclear reactors.

Ignalina NPP comprises two Units with RBMK-1500 reactors. After the Unit 1 of the Ignalina Nuclear Power Plant was shut down in 2004, approximately 1000 fuel assemblies from Unit were available for further reuse in Unit 2. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed in order to implement the process for on-site transportation of Unit 1 fuel for reuse in the Unit 2. The Safety Analysis Report (SAR) was developed to demonstrate that the proposed set of equipment performs all functions and assures the required level of safety for both normal operation and accident conditions. The purpose of this paper is to introduce the content and main results of SAR, focusing attention on the container used to transport spent fuel assemblies from Unit I on Unit 2. In the SAR, the structural integrity, thermal, radiological and nuclearsafety calculations are performed to assess the acceptance of the proposed set of equipment. The safety analysis demonstrated that the proposed nuclear fuel transportation container and other equipment are in compliance with functional, design and regulatory requirements and assure the required safety level. (authors)

Tasks of nuclearsafety assurance for storage and transport of fresh mixed uranium-plutonium fuel of the VVER-1000 reactor are considered in the view of 3 MOX LTAs introduction into the core. The precise code MCU that realizes the Monte Carlo method is used for calculations.

This report is the proceedings of the annual Nuclear Criticality Technology and Safety Project (NCTSP) Workshop held in Monterey, California, on April 16--28, 1993. The NCTSP was sponsored by the Department of Energy and organized by the Los Alamos Critical Experiments Facility. The report is divided into six sections reflecting the sessions outlined on the workshop agenda.

This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Covered are: general-purpose heat source testing and recovery, and safety technology program (biaxial testing, iridium chemistry).

Nuclearsafety-related procedures are rigorous for good reason. Small design mistakes can quickly turn into unwanted failures. Researchers at Oak Ridge National Laboratory worked with COMSOL to define a simulation app that automates the software quality assurance (SQA) verification process and provides results in less than 24 hours.

Safety issues involved in the disposal of nuclear wastes in space as a complement to mined geologic repositories are examined as part of an assessment of the feasibility of nuclear waste disposal in space. General safety guidelines for space disposal developed in the areas of radiation exposure and shielding, containment, accident environments, criticality, post-accident recovery, monitoring systems and isolation are presented for a nuclear waste disposal in space mission employing conventional space technology such as the Space Shuttle. The current reference concept under consideration by NASA and DOE is then examined in detail, with attention given to the waste source and mix, the waste form, waste processing and payload fabrication, shipping casks and ground transport vehicles, launch site operations and facilities, Shuttle-derived launch vehicle, orbit transfer vehicle, orbital operations and space destination, and the system safety aspects of the concept are discussed for each component. It is pointed out that future work remains in the development of an improved basis for the safety guidelines and the determination of the possible benefits and costs of the space disposal option for nuclear wastes.

There is a growing concern over possible adverse effects from medical applications of ionizing radiation. Hospital personnel must be educated in procedures to minimize exposure to themselves and their patients. Basic radiation safety procedures to protect personnel and patients are discussed. Examples of the nuclear medicine policies and procedures used for lymphatic mapping are provided. PMID:10448699

This document includes the Technical requirements which meet the nuclearsafety objectives of the NRC regulations for fuel treatment and storage facilities. These include requirements regarding radiation exposure limits, safety analysis, design and construction. This document also includes administrative requirements which meet the objectives of the major elements of the NRC licensing process. These include formally documented design and safety analysis, independent technical review, and oppportunity for public involvement.

The Fukushima accident raised considerable concern around the globe on the overall safety of nuclear power plants against natural hazard induced risks. Since nuclear power-plants are often located near- or in flood zones from rivers, an important question is whether Nuclear facilities will face increased risk from flooding in the future? IN 2011, the European NuclearSafety Regulators Group (ENSREG) was invited to provide a stress test, as to whether nuclear installations can withstand the consequences of Natural hazards, inclduing flooding. This paper contributes to the findings of ENSREG by demonstrating how global flood risk may increase in the future using a global hydrological model at a 1 x 1 km2 resolution. This information is used to assess the vulnerability of existing and planned nuclear facilities as to whether they (1) are located in flood prone areas (2) are susceptible to an increase in potential flood inundation and (3) are vulnerable to other natural hazards such as earthquake and tsunami. Based on this assessment, a priority ranking can made showing the potentially most vulnerable nuclear power plants to natural hazards, and in particular flood risk.

The purpose of this analysis, which was conducted for the US Nuclear Regulatory Commission (NRC), was to test whether an empirical connection exists between human performance and nuclear power plant safety culture. This was accomplished through analyzing the relationship between a measure of human performance and a plant’s Safety Conscious Work Environment (SCWE). SCWE is an important component of safety culture the NRC has developed, but it is not synonymous with it. SCWE is an environment in which employees are encouraged to raise safety concerns both to their own management and to the NRC without fear of harassment, intimidation, retaliation, or discrimination. Because the relationship between human performance and allegations is intuitively reciprocal and both relationship directions need exploration, two series of analyses were performed. First, human performance data could be indicative of safety culture, so regression analyses were performed using human performance data to predict SCWE. It also is likely that safety culture contributes to human performance issues at a plant, so a second set of regressions were performed using allegations to predict HFIS results.

One of the challenges in educating our next generation of nuclearsafety engineers is the limitation of opportunities to receive significant experience or hands-on training prior to graduation. Such training is generally restricted to on-the-job-training before this new engineering workforce can adequately provide assessment of nuclear systems and establish safety guidelines. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) can provide students and young professionals the opportunity to gain experience and enhance critical engineering skills. The ICSBEP and IRPhEP publish annual handbooks that contain evaluations of experiments along with summarized experimental data and peer-reviewed benchmark specifications to support the validation of neutronics codes, nuclear cross-section data, and the validation of reactor designs. Participation in the benchmark process not only benefits those who use these Handbooks within the international community, but provides the individual with opportunities for professional development, networking with an international community of experts, and valuable experience to be used in future employment. Traditionally students have participated in benchmarking activities via internships at national laboratories, universities, or companies involved with the ICSBEP and IRPhEP programs. Additional programs have been developed to facilitate the nuclear education of students while participating in the benchmark projects. These programs include coordination with the Center for Space Nuclear Research (CSNR) Next Degree Program, the Collaboration with the Department of Energy Idaho Operations Office to train nuclear and criticality safety engineers, and student evaluations as the basis for their Master's thesis in nuclear engineering.

In order to maintain the operating plant safety at current safety standards, Periodic Safety Review (PSR) is legislated in 2001 in Korea as a 10-year-basis safety evaluation process. One of the eleven topics addressed in the PSR is safety analysis, in which the compliance of plants' safety analyses with current standards on initiating events and scope, methods and assumptions are evaluated. This paper describes the methods and results of the PSR on safety analysis of the Kori nuclear units 3,4, 3-Loop pressurized water reactors in operation from 1984. The review areas are design-basis, beyond-design basis and severe accidents. Evaluation of design basis accidents (DBA) in the Kori units 3,4 Final Safety Analysis Report (FSAR) issued a lack of consideration of loss of offsite power (LOOP) in some DBAs. To resolve this, electric power system stability analysis has been performed to show that sufficient time delay of reactor coolant pump trip after LOOP makes the current FSAR DBA analyses still valid with additional assumption of LOOP as an initiating event. Also, to get confidence on defense-in-depth safety, thermal hydraulic analyses are performed for beyond-design basis and severe accidents. A typical high-pressure scenario, total loss of feedwater event, is selected and analyzed by using RELAP5/MOD3 and MAAP4 computer codes for recovered and un-recovered cases, respectively. It is shown that using safety-grade pressurizer relief valves, the two cases meet the criteria that are applied to the new Korean Standard Nuclear Plants (KSNP) in Korea. It is thus concluded that the Kori units 3,4 have good design capabilities to prevent and mitigate broad spectrum of reactor accidents from design basis to severe accidents. (authors)

... information regarding the quality and safety of health care delivery. The Patient Safety and Quality... is to conduct activities to improve patient safety and the quality of health care delivery. HHS... HUMAN SERVICES Agency for Healthcare Research and Quality Patient Safety Organizations:...

Analysis of the capabilities of the State Corporation Rosatom enterprises on the development and production of diagnostic and therapeutic equipment for nuclear medicine centers is presented. Prospects of the development of accelerator equipment for the production of a wide range of radioisotope products are shown, and the trends of its development are determined. A comparative analysis of the technical parameters of domestic tomographs and devices for brachytherapy with foreign counterparts is given.

Analysis of the capabilities of the State Corporation Rosatom enterprises on the development and production of diagnostic and therapeutic equipment for nuclear medicine centers is presented. Prospects of the development of accelerator equipment for the production of a wide range of radioisotope products are shown, and the trends of its development are determined. A comparative analysis of the technical parameters of domestic tomographs and devices for brachytherapy with foreign counterparts is given.

Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

Advanced control rooms (ACR's) will utilize human-system interface (HSI) technologies that may have significant implications for plant safety in that they will affect the operator's overall role and means of interacting with the system. The Nuclear Regulatory Commission (NRC) reviews the human factors engineering (HFE) aspects of HSI's to ensure that they are designed to good HFE principles and support performance and reliability in order to protect public health and safety. However, the only available NRC guidance was developed more than ten years ago, and does not adequately address the human performance issues and technology changes associated with ACR's. Accordingly, a new approach to ACR safety reviews was developed based upon the concept of 'convergent validity'. This approach to ACR safety reviews is described.

This document contains summaries of the most of the papers presented at the 1994 Nuclear Criticality Technology Safety Project (NCTSP) meeting, which was held May 10 and 11 at Williamsburg, Va. The meeting was broken up into seven sessions, which covered the following topics: (1) Validation and Application of Calculations; (2) Relevant Experiments for Criticality Safety; (3) Experimental Facilities and Capabilities; (4) Rad-Waste and Weapons Disassembly; (5) Criticality Safety Software and Development; (6) Criticality Safety Studies at Universities; and (7) Training. The minutes and list of participants of the Critical Experiment Needs Identification Workgroup meeting, which was held on May 9 at the same venue, has been included as an appendix. A second appendix contains the names and addresses of all NCTSP meeting participants. Separate abstracts have been indexed to the database for contributions to this proceedings.

The Interagency NuclearSafety Review Panel (INSRP) has provided reviews of all nuclear powered spacecraft launched by the United States. The two most recent launches were Ulysses in 1990 and Galileo in 1989. One reactor was launched in 1965 (SNAP-10A). All other U.S. space missions have utilized radioisotopic thermoelectric generators (RTGs). There are several missions in the next few years that are to be nuclear powered, including one that would utilize the Topaz II reactor purchased from Russia. INSRP must realign itself to perform parallel safety assessments of a reactor powered space mission, which has not been done in about thirty years, and RTG powered missions.

The Interagency NuclearSafety Review Panel (INSRP) has provided reviews of all nuclear powered spacecraft launched by the United States. The two most recent launches were Ulysses in 1990 and Galileo in 1989. One reactor was launched in 1965 (SNAP-10A). All other U.S. space missions have utilized radioisotopic thermoelectric generators (RTGs). There are several missions in the next few years that are to be nuclear powered, including one that would utilize the Topaz II reactor purchased from Russia. INSRP must realign itself to perform parallel safety assessments of a reactor powered space mission, which has not been done in about thirty years, and RTG powered missions.

While issues like dry storage, low-level waste storage, radiation, and license extension are projects with a technical perspective that need to be planned and executed at nuclear power plants, more difficult is the political perspective-gaining public acceptance to allow these projects to proceed. And public perception is predicated on the way plant neighbors and community members understand, accept, and trust the plants. Community educators are a key audience. Annually, U.S. information centers host about one million visitors; roughly half of them are school children who will soon join the ranks of voters, taxpayers, utility customers, and employees. Programs for educators and their classes vary from tours of centers that include computer games and video programs on energy-related topics to audio-visual presentations by center personnel. Some facilities have environmental activities such as hatcheries or nature trails, while others offer plant tours to specific age groups.

This paper outlines the use of a Failure Modes and Effects Analysis for the safety assessment of a robotic system being developed at Sandia National Laboratories. The robotic system, The Weigh and Leak Check System, is to replace a manual process at the Department of Energy facility at Pantex by which nuclear material is inspected for weight and leakage. Failure Modes and Effects Analyses were completed for the robotics process to ensure that safety goals for the system had been meet. These analyses showed that the risks to people and the internal and external environment were acceptable.

Digital instrumentation and control systems have the potential of offering significant benefits over traditional analog systems in Nuclear Power Plant safety systems, but there are also significant difficulties in qualifying digital systems to the satisfaction of regulators. Digital systems differ in fundamental ways from analog systems. New methods for safety qualification, which take these differences into account, would ease the regulatory cost and promote use of digital systems. This paper offers a possible method for assisting in the analysis of digital system software, as one step in an improved qualification process.

In the aftermath of Fukushima natural-technological disaster the global opinion on nuclear energy divided even deeper. While Germany, Italy and the USA are currently reevaluating their previous plans on nuclear growth, many states are committed to expand nuclear energy output. In China and France, where the industry is widely supported by policymakers, there is little talk about abandoning further development of nuclear energy. Moreover, China displays the most remarkable pace of nuclear development in the world: it is responsible for 40% of worldwide reactors under construction, and aims at least to quadruple its nuclear capacity by 2020. In these states the consequences of Fukushima natural-technological accident will probably result in safety checks and advancement of new reactor technologies. Thus, China is buying newer reactor design from the USA which relies on "passive safety systems". It means that emergency power generators, crucial for reactor cooling in case of an accident, won't depend on electricity, so that tsunami won't disable them like it happened in the case of Fukushima. Nuclear energy managed to draw lessons from previous nuclear accidents where technological and human factors played crucial role. But the Fukushima lesson shows that the natural hazards, nevertheless, were undervalued. Though the ongoing technological advancements make it possible to increase the safety of nuclear power plants with consideration of natural risks, it is not just a question of technology improvement. A necessary action that must be taken is the reevaluation of the character and sources of the potential hazards which natural disasters can bring to nuclear industry. One of the examples is a devastating impact of more than one natural disaster happening at the same time. This subject, in fact, was not taken into account before, while it must be a significant point in planning sites for new nuclear power plants. Another important lesson unveiled is that world nuclear

... National Resource Center for Health and Safety in Child Care and Early Education (NRC) at the University of ... State of Healthy Weight (annual national assessments of child care regulations relative to healthy weight-specific CFOC3 standards ...

In this paper the authors identify factors which influence the safety philosophy used in the US commercial/industrial sector and compare them against those factors which influence nuclear weapons safety. Commercial/industrial safety is guided by private and public safety standards. Generally, private safety standards tend to emphasize product reliability issues while public (i.e., government) safety standards tend to emphasize human factors issues. Safety in the nuclear weapons arena is driven by federal requirements and memoranda of understanding (MOUs) between the Departments of Defense and Energy. Safety is achieved through passive design features integrated into the nuclear weapon. Though the common strand between commercial/industrial and nuclear weapons safety is the minimization of risk posed to the general population (i.e., public safety), the authors found that each sector tends to employ a different safety approach to view and resolve high-consequence safety issues.

Staff members, physicians, nurse practitioners, and physician assistants from a sample of hemodialysis facilities in Network 6 (North Carolina, South Carolina, and Georgia) and Network 11 (Michigan, Minnesota, North Dakota, South Dakota, and Wisconsin) completed a 10-item assessment with modified questions from the Hospital Survey on Patient Safety Culture, with an emphasis on safety culture related to vascular access infections. A composite score was constructed, which was the average of the percent-positive scores of the items. Overall, scores were high, indicating a positive patient safety culture. Composite scores varied by role type, with nurses, patient care technicians, and other technicians reporting the lowest composite scores. Network 6 participants reported higher scores on two of the survey items. Fewer staff within a facility were associated with higher composite scores. PMID:27254967

Nuclear power has been considered for space applications since the 1960s. Between 1955 and 1972 the US built and tested over twenty nuclear reactors/ rocket-engines in the Rover/NERVA programs. However, changes in environmental laws may make the redevelopment of the nuclear rocket more difficult. Recent advances in fuel fabrication and testing options indicate that a nuclear rocket with a fuel form significantly different from NERVA may be needed to ensure public support. The Center for Space Nuclear Research (CSNR) is pursuing development of tungsten based fuels for use in a NTR, for a surface power reactor, and to encapsulate radioisotope power sources. The CSNR Summer Fellows program has investigated the feasibility of several missions enabled by the NTR. The potential mission benefits of a nuclear rocket, historical achievements of the previous programs, and recent investigations into alternatives in design and materials for future systems will be discussed.

Under a commission from the Office of Nuclear Regulatory Research, NRC, a study was performed by a team of analysts relative to licensing practices and the role of security as they relate to safeguards during safety-related emergencies (SREs). Methodology included a literature search, site visits to representative nuclear reactors and analysis of the regulatory and licensee planning bases. Problems relating to security actions during SREs were examined primarily in the following areas: organization for response, planning, training and qualification, equipment, procedures employed, facilities, and preparation for safeguards against sabotage during an SRE. Recommendations were made as to how improvements could be made in the regulatory approach, and in licensee planning and procedural mechanisms as they relate to the subject matter under examination. The results of the study also had implications for the safety/safeguards interface problem currently under review by the NRC.

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

The Savannah River Site (SRS) High Level Nuclear Waste will be vitrified in the Defense Waste Processing Facility (DWPF) for long term storage and disposal. This is a preliminary safety evaluation for the Melt Cell of the DWPF vitrification process for Batch 1 waste. This evaluation demonstrates that the material in the Melt cell remains subcritical for the contents of Batch 1 which contains uranium with less than 1% by weight U-235.

The SP-100 Nuclear Assembly Test (NAT) article is charged with the validation of such a fast-spectrum, liquid metal-cooled reactor's integrated reactor, control, and shield performance characteristics in a spacelike environment. The NAT facility is designed for safe assembly, operation, disassembly, and disposal. Attention is presently given to the selection and classification of postulated accidents, safety design criteria, reactor trip parameters, and the structure of accident analyses.

This paper focuses on work practices that may contribute to fatigue-induced performance decrements in the commercial nuclear power industry. Specifically, the amount of overtime worked by operations, technical, and maintenance personnel and the 12-h operator shift schedule are studied. Although overtime for all three job categories was fairly high at a number of plants, the analyses detected a clear statistical relationship only between operations overtime and plant safety performance. The results for the 12-h operator shift schedule were ambiguous. Although the 12-h operator shift schedule was correlated with operator error, it was not significantly related to the other five safety indicators. This research suggests that at least one of the existing work practices--the amount of operator overtime worked at some plants--represents a safety concern in this industry; however, further research is required before any definitive conclusions can be drawn. PMID:8070790

The US Nuclear Regulatory Commission (NRC) is evaluating the possibility of instituting a data gathering system for identifying and quantifying the factors that contribute to the occurrence of significant safety problems involving humans in nuclear power plants. This report presents the results of a brief (6 months) study of the feasibility of developing a voluntary, nonpunitive Nuclear Power Safety Reporting System (NPSRS). Reports collected by the system would be used to create a data base for documenting, analyzing and assessing the significance of the incidents. Results of The Aerospace Corporation study are presented in two volumes. This document, Volume I, contains a summary of an assessment of the Aviation Safety Reporting System (ASRS). The FAA-sponsored, NASA-managed ASRS was found to be successful, relatively low in cost, generally acceptable to all facets of the aviation community, and the source of much useful data and valuable reports on human factor problems in the nation's airways. Several significant ASRS features were found to be pertinent and applicable for adoption into a NPSRS.

The French nuclearsafety authority is responsible for the control of radiation protection in radiotherapy since 2002. Controls are based on the public health and the labour codes and on the procedures defined by the controlled health care facility for its quality and safety management system according to ASN decision No. 2008-DC-0103. Inspectors verify the adequacy of the quality and safety management procedures and their implementation, and select process steps on the basis of feedback from events notified to ASN. Topics of the inspection are communicated to the facility at the launch of a campaign, which enables them to anticipate the inspectors' expectations. In cases where they are not physicians, inspectors are not allowed to access information covered by medical confidentiality. The consulted documents must therefore be expunged of any patient-identifying information. Exchanges before the inspection are intended to facilitate the provision of documents that may be consulted. Finally, exchange slots between inspectors and the local professionals must be organized. Based on improvements achieved by the health care centres and on recommendations from a joint working group of radiotherapy professionals and the nuclearsafety authority, changes will be made in the control procedure that will be implemented when developing the inspection program for 2016-2019. PMID:26321685

The U.S. Department of Energy, Richland Operations Office (RL) is responsible for 30 hazard category 2 and 3 nuclear facilities that are operated by its prime contractors, Fluor Hanford, Incorporated (FHI), Bechtel Hanford, Incorporated (BHI) and Pacific Northwest National Laboratory (PNNL). The publication of Title 10, Code of Federal Regulations, Part 830, Subpart B, Safely Basis Requirements (the Rule) in January 2001 requires that the Documented Safety Analyses (DSA) for these facilities be reviewed against the requirements of the Rule. Those DSAs that do not meet the requirements must either be upgraded to satisfy the Rule, or an exemption must be obtained. RL and its prime contractors have developed a NuclearSafety Strategy that provides a comprehensive approach for supporting RL's efforts to meet its long-term objectives for hazard category 2 and 3 facilities while also meeting the requirements of the Rule. This approach will result in a reduction of the total number of safety basis documents that must be developed and maintained to support the remaining mission and closure of the Hanford Site and ensure that the documentation that must be developed will support: Compliance with the Rule; A ''Just-In-Time'' approach to development of Rule-compliant safety bases supported by temporary exemptions; and Consolidation of safety basis documents that support multiple facilities with a common mission (e.g. decontamination, decommissioning and demolition [DD&D], waste management, surveillance and maintenance). This strategy provides a clear path to transition the safety bases for the various Hanford facilities from support of operation and stabilization missions through DD&D to accelerate closure. This ''Just-In-Time'' Strategy can also be tailored for other DOE Sites, creating the potential for large cost savings and schedule reductions throughout the DOE complex.

The purpose of the crash safety program is to support development of the technology to define and demonstrate new structural concepts for improved crash safety and occupant survivability in general aviation aircraft. The program involves three basic areas of research: full-scale crash simulation testing, nonlinear structural analyses necessary to predict failure modes and collapse mechanisms of the vehicle, and evaluation of energy absorption concepts for specific component design. Both analytical and experimental methods are being used to develop expertise in these areas. Analyses include both simplified procedures for estimating energy absorption capabilities and more complex computer programs for analysis of general airframe response. Full-scale tests of typical structures as well as tests on structural components are being used to verify the analyses and to demonstrate improved design concepts.

George Washington University (GW) has one of the largest university-based nuclear-physics groups in the nation. Many of the current and future projects are geared to Thomas Jefferson National Accelerator Facility (JLab) at Newport News, VA. JLab is the world's premier electron accelerator for nuclear physics, and GW is one of the charter members of the governing body of JLab, the Southeastern Universities Research Association (SURA). The George Washington Data Analysis Center (DAC) was created in 1998 by an agreement among the Department of Energy, Jefferson Lab, and the GW Center for Nuclear Studies.The activities of the DAC fall into four distinct categories: 1) Performing partial-wave analyses of fundamental two- and three-body reactions; 2) Maintenance of databases associated with these reactions; 3) Development of software to disseminate DAC results (as well as the results of competing model-independent analyses and potential approaches); and 4) Phenomenological and theoretical investigations which bridge the gap between theory and experiment; in particular, the extraction of N* and D * hadronic and electromagnetic couplings. Partial Wave Analyses (and the associated databases) available at GW are: Pion-Nucleon, Kaon-Nucleon, Nucleon-Nucleon, Pion Photoproduction, Pion Electroproduction, Kaon Photoproduction, Eta Photoproduction, Eta-Prime Photoproduction, Pion-Deuteron (elastic), and Pion-Deuteron to Proton+Proton. [Taken from http://www.gwu.edu/~ndl/dac.htm">http://www.gwu.edu/~ndl/dac.htm

An Interagency NuclearSafety Policy Working Group (NSPWG) was chartered to recommend nuclearsafety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top- level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of Safety Functional Requirements. In addition the NSPWG reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. Safety requirements were developed for reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, and safeguards. Guidelines were recommended for risk/reliability, operational safety, flight trajectory and mission abort, space debris and meteoroids, and ground test safety. In this paper the specific requirements and guidelines will be discussed.

An Interagency NuclearSafety Policy Working Group (NSPWG) was chartered to recommend nuclearsafety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top- level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of Safety Functional Requirements. In addition the NSPWG reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. Safety requirements were developed for reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, and safeguards. Guidelines were recommended for risk/reliability, operational safety, flight trajectory and mission abort, space debris and meteoroids, and ground test safety. In this paper the specific requirements and guidelines will be discussed.

An interagency NuclearSafety Policy Working Group (NSPWG) was chartered to recommend nuclearsafety policy, requirements, and guidelines for the Space Exploration Initiative nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of Safety Functional Requirements. In addition, the NSPWG reviewed safety issues for nuclear propulsion and recommended top level safety requirements and guidelines to address these issues. Safety topics include reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, safeguards, risk/reliability, operational safety, ground testing, and other considerations. In this paper the emphasis is placed on the safety policy and the issues and considerations that are addressed by the NSPWG recommendations.

An interagency NuclearSafety Policy Working Group (NSPWG) was chartered to recommend nuclearsafety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top-level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of safety functional requirements. In addition, the NSPWG reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. Safety requirements were developed for reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, and safeguards. Guidelines were recommended for risk/reliability, operational safety, flight trajectory and mission abort, space debris and meteoroids, and ground test safety. In this paper the specific requirements and guidelines will be discussed.

An interagency NuclearSafety Policy Working Group (NSPWG) was chartered to recommend nuclearsafety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top-level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of safety functional requirements. In addition, the NSPWG reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. Safety requirements were developed for reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, and safeguards. Guidelines were recommended for risk/reliability, operational safety, flight trajectory and mission abort, space debris and meteoroids, and ground test safety. In this paper the specific requirements and guidelines will be discussed. 9 refs.

This manual is a tool to provide information to all responders and emergency planners and is suggested as a starting point for all organizations that provide personnel/assets for radiological emergency response. It defines the safety requirements for the protection of all emergency responders. The intent is to comply with appropriate regulations or provide an equal level of protection when the situation makes it necessary to deviate. In the event a situation arises which is not addressed in the manual, an appropriate management-level expert will define alternate requirements based on the specifics of the emergency situation. This manual is not intended to pertain to the general public.

The DOE Office of Enforcement expects LLNL to 'implement comprehensive management and independent assessments that are effective in identifying deficiencies and broader problems in safety and security programs, as well as opportunities for continuous improvement within the organization' and to 'regularly perform assessments to evaluate implementation of the contractor's processes for screening and internal reporting.' LLNL has a self-assessment program, described in ES&H Manual Document 4.1, that includes line, management and independent assessments. LLNL also has in place a process to identify and report deficiencies of nuclear, worker safety and health and security requirements. In addition, the DOE Office of Enforcement expects LLNL to evaluate 'issues management databases to identify adverse trends, dominant problem areas, and potential repetitive events or conditions' (page 14, DOE Enforcement Process Overview, December 2007). LLNL requires that all worker safety and health and nuclearsafety noncompliances be tracked as 'deficiencies' in the LLNL Issues Tracking System (ITS). Data from the ITS are analyzed for worker safety and health (WSH) and nuclearsafety noncompliances that may meet the threshold for reporting to the DOE Noncompliance Tracking System (NTS). This report meets the expectations defined by the DOE Office of Enforcement to review the assessments conducted by LLNL, analyze the issues and noncompliances found in these assessments, and evaluate the data in the ITS database to identify adverse trends, dominant problem areas, and potential repetitive events or conditions. The report attempts to answer three questions: (1) Is LLNL evaluating its programs and state of compliance? (2) What is LLNL finding? (3) Is LLNL appropriately managing what it finds? The analysis in this report focuses on data from the first quarter of 2008 (January through March). This quarter is analyzed within the context of information identified in previous quarters to

A Reliability Program (RP) model based on proven reliability techniques used in other high technology industries is being formulated for potential application in the nuclear power industry. Research findings are discussed. The reliability methods employed under NASA and military direction, commercial airline and related FAA programs were surveyed with several reliability concepts (e.g., quantitative reliability goals, reliability centered maintenance) appearing to be directly transferable. Other tasks in the RP development effort involved the benchmarking and evaluation of the existing nuclear regulations and practices relevant to safety/reliability integration. A review of current risk-dominant issues was also conducted using results from existing probabilistic risk assessment studies. The ongoing RP development tasks have concentrated on defining a RP for the operating phase of a nuclear plant's lifecycle. The RP approach incorporates safety systems risk/reliability analysis and performance monitoring activities with dedicated tasks that integrate these activities with operating, surveillance, and maintenance of the plant. The detection, root-cause evaluation and before-the-fact correction of incipient or actual systems failures as a mechanism for maintaining plant safety is a major objective of the RP.

The Columbia Accident Investigation Board (CAIB) determined that organizational and management issues were significant contributors to the loss of Space Shuttle Columbia. In addition, the CAIB observed similarities between the organizational and management climate that preceded the Challenger accident and the climate that preceded the Columbia accident. To prevent recurrence of adverse organizational and management climates, effective implementation of the system safety function is suggested. Attributes of an effective system safety program are presented. The Marshall Space Flight Center (MSFC) system safety program is analyzed using the attributes. Conclusions and recommendations for improving the MSFC system safety program are offered in this case study.

The Environmental Assessment (EA) of the Center for Nuclear Medicine Research in Alzheimer`s Disease (CNMR) at the Health Sciences Center, at West Virginia University in Morgantown, West Virginia for the construction and operation was prepared by DOE. The EA documents analysis of the environmental and socioeconomic impacts that might occur as a result of these actions, and characterizes potential impacts on the environment. In the EA, DOE presents its evaluation of potential impacts of construction and operation of the CNMR on health and safety of both workers and the public, as well as on the external environment. Construction impacts include the effects of erosion, waste disposal, air emissions, noise, and construction traffic and parking. Operational impacts include the effects of waste generation (domestic, sanitary, hazardous, medical/biological, radioactive and mixed wastes), radiation exposures, air emissions (radioactive, criteria, and air toxics), noise, and new workers. No sensitive resources (wetlands, special sources of groundwater, protected species) exist in the area of project effect.

The Office of Science, Basic Energy Sciences, has funded the INL as one of the Energy Frontier Research Centers in the area of material science of nuclear fuels. This document is the required annual report to the Office of Science that outlines the accomplishments for the period of May 2010 through April 2011. The aim of the Center for Material Science of Nuclear Fuels (CMSNF) is to establish the foundation for predictive understanding of the effects of irradiation-induced defects on thermal transport in oxide nuclear fuels. The science driver of the center’s investigation is to understand how complex defect and microstructures affect phonon mediated thermal transport in UO2, and achieve this understanding for the particular case of irradiation-induced defects and microstructures. The center’s research thus includes modeling and measurement of thermal transport in oxide fuels with different levels of impurities, lattice disorder and irradiation-induced microstructure, as well as theoretical and experimental investigation of the evolution of disorder, stoichiometry and microstructure in nuclear fuel under irradiation. With the premise that thermal transport in irradiated UO2 is a phonon-mediated energy transport process in a crystalline material with defects and microstructure, a step-by-step approach will be utilized to understand the effects of types of defects and microstructures on the collective phonon dynamics in irradiated UO2. Our efforts under the thermal transport thrust involved both measurement of diffusive phonon transport (an approach that integrates over the entire phonon spectrum) and spectroscopic measurements of phonon attenuation/lifetime and phonon dispersion. Our distinct experimental efforts dovetail with our modeling effort involving atomistic simulation of phonon transport and prediction of lattice thermal conductivity using the Boltzmann transport framework.

The Gaerttner LINAC Center at RPI uses a 60 MeV electron linear accelerator to produce short pulses of neutrons with duration of 5-5000 ns. The main research thrust at the Center is nuclear data for nuclear reactors and criticality safety applications. The Center includes several setups for time-of-flight measurements including neutron transmission, capture and scattering detectors, and a lead slowing-down spectrometer. Experiments were designed to produce neutron interaction cross sections that cover the energy range of 0.01 eV to 20 MeV. Recently added experiments include: setups for keV and fast neutron transmission, a C6D6 detector array for keV neutron capture measurements, and a fast neutron scattering system. Results discussed here include fast neutron scattering and angular distributions for natFe, iron capture measurements for incident neutrons from 1 keV to 2 MeV, fast neutron transmission through W and H2O samples, and keV transmission through Mo isotopes.

Washington Safety Management Solutions (WSMS) provides criticality safety services to Washington Savannah River Company (WSRC) at the Savannah River Site. One activity at SRS is the Container Surveillance and Storage Capability (CSSC) Project, which will perform surveillances on 3013 containers (hereafter referred to as 3013s) to verify that they meet the Department of Energy (DOE) Standard (STD) 3013 for plutonium storage. The project will handle quantities of material that are greater than ANS/ANSI-8.1 single parameter mass limits, and thus required a Nuclear Criticality Safety Evaluation (NCSE). The WSMS methodology for conducting an NCSE is outlined in the WSMS methods manual. The WSMS methods manual currently follows the requirements of DOE-O-420.1B, DOE-STD-3007-2007, and the Washington Savannah River Company (WSRC) SCD-3 manual. DOE-STD-3007-2007 describes how a NCSE should be performed, while DOE-O-420.1B outlines the requirements for a Criticality Safety Program (CSP). The WSRC SCD-3 manual implements DOE requirements and ANS standards. NCSEs do not address the Nuclear Criticality Safety (NCS) of non-reactor nuclear facilities that may be affected by overt or covert activities of sabotage, espionage, terrorism or other security malevolence. Events which are beyond the Design Basis Accidents (DBAs) are outside the scope of a double contingency analysis.

The March 2011 accident at the Fukushima Daiichi nuclear power plant has revealed serious vulnerabilities in the design, operation and regulation of nuclear power plants. While some aspects of the accident were plant- and site-specific, others have implications that are broadly applicable to the current generation of nuclear plants in operation around the world. Although many of the details of the accident progression and public health consequences are still unclear, there are a number of lessons that can already be drawn. The accident demonstrated the need at nuclear plants for robust, highly reliable backup power sources capable of functioning for many days in the event of a complete loss of primary off-site and on-site electrical power. It highlighted the importance of detailed planning for severe accident management that realistically evaluates the capabilities of personnel to carry out mitigation operations under extremely hazardous conditions. It showed how emergency plans rooted in the assumption that only one reactor at a multi-unit site would be likely to experience a crisis fail miserably in the event of an accident affecting multiple reactor units simultaneously. It revealed that alternate water injection following a severe accident could be needed for weeks or months, generating large volumes of contaminated water that must be contained. And it reinforced the grim lesson of Chernobyl: that a nuclear reactor accident could lead to widespread radioactive contamination with profound implications for public health, the economy and the environment. While many nations have re-examined their policies regarding nuclear power safety in the months following the accident, it remains to be seen to what extent the world will take the lessons of Fukushima seriously and make meaningful changes in time to avert another, and potentially even worse, nuclear catastrophe.

Robotic systems are being developed by the Intelligent Systems and Robotics Center at Sandia National Laboratories to perform automated handling tasks with radioactive weapon parts. These systems will reduce the occupational radiation exposure to workers by automating operations that are currently performed manually. The robotic systems at Sandia incorporate several levels of mechanical, electrical, and software safety for handling hazardous materials. For example, tooling used by the robot to handle radioactive parts has been designed with mechanical features that allow the robot to release its payload only at designated locations in the robotic workspace. In addition, software processes check for expected and unexpected situations throughout the operations. Incorporation of features such as these provides multiple levels of safety for handling hazardous or valuable payloads with automated intelligent systems.

Objectives Safety Net Health Centers (SNHCs), which include Federally Qualified Health Centers (FQHCs) provide primary care for underserved, minority and low income patients. SNHCs across the country are in the process of adopting the Patient Centered Medical Home (PCMH) model, based on promising early implementation data from demonstration projects. However, previous demonstration projects have not focused on the safety net and we know little about PCMH transformation in SNHCs. Design This qualitative study characterizes early PCMH adoption experiences at SNHCs. Setting and Participants We interviewed 98 staff,(administrators, providers, and clinical staff) at 20 of 65 SNHCs, from five states, who were participating in the first of a five-year PCMH collaborative, the Safety Net Medical Home Initiative. Main Measures We conducted 30-45 minute, semi-structured telephone interviews. Interview questions addressed benefits anticipated, obstacles encountered, and lessons learned in transition to PCMH. Results Anticipated benefits for participating in the PCMH included improved staff satisfaction and patient care and outcomes. Obstacles included staff resistance and lack of financial support for PCMH functions. Lessons learned included involving a range of staff, anticipating resistance, and using data as frequent feedback. Conclusions SNHCs encounter unique challenges to PCMH implementation, including staff turnover and providing care for patients with complex needs. Staff resistance and turnover may be ameliorated through improved healthcare delivery strategies associated with the PCMH. Creating predictable and continuous funding streams may be more fundamental challenges to PCMH transformation. PMID:23914423

The results of a survey of organizational culture at a nuclear power plant are summarized and compared with those of a similar survey which has been described in the literature on ''high-reliability organizations''. A general-purpose cultural inventory showed a profile of organizational style similar to that reported in the literature; the factor structure for the styles was also similar to that of the plant previously described. A specialized scale designed to measure ''safety culture'' did not distinguished among groups within the organization that would be expected to differ.

The Three Mile Island nuclear power plant accident has resulted in redirection of reactor safety research priorities. The small release to the environment of radioactive iodine-13 to 17 curies in a total radioactivity release of 2.4 million to 13 million curies-has led to a new emphasis on the physical chemistry of fission product behavior in accidents; the fact that the nuclear core was severely damaged but did not melt down has opened a new accident regime-that of the degraded core; the role of the operators in the progression and severity of the accident has shifted emphasis from equipment reliability to human reliability. As research progresses in these areas, the technical base for regulation and risk analysis will change substantially. PMID:17736229

To evaluate the hygienic risk involved in using baculoviruses for insect pest control, safety studies are required. Pigs were chosen as representative test animals of commercial and agricultural importance. The tests were aimed at detecting virus propagation, immune reactions, and signs of acute infection (changes in body temperature and hematology profile, swelling of lymph nodes). Four of five animals inoculated with nuclear polyhedrosis virus showed a slight temperature rise at day 2 postinfection. After day 4 postinfection, no differences between infected animals and controls were observed. In the bioassay, virus activity could be recovered from fecal samples; however, no activity was found in organ extracts. The data did not indicate hygienic risks involved in the application of nuclear polyhedrosis virus, especially that from Mamestra brassicae, in biological pest control. PMID:6344789

To evaluate the hygienic risk involved in using baculoviruses for insect pest control, safety studies are required. Pigs were chosen as representative test animals of commercial and agricultural importance. The tests were aimed at detecting virus propagation, immune reactions, and signs of acute infection (changes in body temperature and hematology profile, swelling of lymph nodes). Four of five animals inoculated with nuclear polyhedrosis virus showed a slight temperature rise at day 2 postinfection. After day 4 postinfection, no differences between infected animals and controls were observed. In the bioassay, virus activity could be recovered from fecal samples; however, no activity was found in organ extracts. The data did not indicate hygienic risks involved in the application of nuclear polyhedrosis virus, especially that from Mamestra brassicae, in biological pest control. PMID:6344789

The Institute for Safe Medication Practices encourages pharmacists to assist in preventing medication misuse. The purpose of this article is to discuss a medication-safety education session conducted by a pharmacy professor, the faculty advisor to the American Society of Consultant Pharmacists university student chapter and students in a pharmacovigilance rotation, which was conducted at a local senior center. The author attended a train-the-trainer Webinar and then educated the pharmacy students. Participants at the senior center were taught about poison prevention, drug interactions, and appropriate drug disposal through an interactive format. We plan to continue the medication safety program at the senior center as a longitudinal project to promote patient safety. Pharmacists should be encouraged to play an active role in community outreach programs. PMID:24129222

This presentation provides an update to the author's standards report provided at the ICNC-2007 meeting. It includes a discussion about the difference between, and the value of participating in, the development of international 'consensus' standards as opposed to nonconsensus standards. Standards are developed for a myriad of reasons. Generally, standards represent an agreed upon, repeatable way of doing something as defined by an individual or group of people. They come in various types. Examples include personal, family, business, industrial, commercial, and regulatory such as military, community, state, federal, and international standards. Typically, national and international 'consensus' standards are developed by individuals and organizations of diverse backgrounds representing the subject matter users and developers of a service or product and other interested parties or organizations. Within the International Organization for Standardization (ISO), Technical Committee 85 (TC85) on nuclear energy, Subcommittee 5 (SC5) on nuclear fuel technology, there is a Working Group 8 (WG8) on standardization of calculations, procedures, and practices related to criticality safety. WG8 has developed, and is developing, ISO standards within the category of nuclear criticality safety of fissionable materials outside of reactors (i.e., nonreactor fissionable material nuclear fuel cycle facilities). Since the presentation of the ICNC-2007 report, WG8 has issued three new finalized international standards and is developing two more new standards. Nearly all elements of the published WG8 ISO standards have been incorporated into IAEA nonconsensus guides and standards. The progression of consensus standards development among international partners in a collegial environment establishes a synergy of different concepts that broadens the perspectives of the members. This breadth of perspectives benefits the working group members in their considerations of consensus standards

One year after Japan suffered a devastating magnitude 9.0 earthquake and the resulting tsunami and nuclear disaster, the U.S. Nuclear Regulatory Commission (NRC) has taken a number of measures to try to ensure that nuclear plants in the United States are safe from natural hazards. At a U.S. Senate hearing on 15 March, NRC chair Gregory Jaczko announced that the commission had issued three key orders and several requests for information on 12 March that plant licensees must follow, and that NRC also plans to take additional actions. However, the commission is not moving quickly enough in some areas, such as ensuring that all plants are safe from seismic hazards, including those in areas with low seismic activity, according to Jaczko's testimony before the Senate Committee on Environment and Public Works (EPW) and the Subcommittee on Clean Air and NuclearSafety. The 12 March orders require licensees to have strategies to maintain or restore core cooling, containment, and spent-fuel pool cooling capabilities "following a beyond-design-basis extreme natural event" and have a reliable indication of the water level in spent-fuel storage pools.

Covariance evaluations at the China Nuclear Data Center focus on the cross sections of structural materials and actinides in the fast neutron energy range. In addition to the well-known Least-squares approach, a method based on the analysis of the sources of experimental uncertainties is especially introduced to generate a covariance matrix for a particular reaction for which multiple measurements are available. The scheme of the covariance evaluation flow is presented, and an example of n+{sup 90}Zr is given to illuminate the whole procedure. It is proven that the accuracy of measurements can be properly incorporated into the covariance and the long-standing small uncertainty problem can be avoided.

The Embrittlement Data Base (EDB) is a comprehensive collection of data from surveillance capsules of U.S. commercial nuclear power reactors and from experiments in material test reactors. The collected data are contained in either the Power Reactor Embrittlement Data Base (PR-EDB) or the Test Reactor Embrittlement Data Base (TR-EDB). The EDB work includes verification of the quality of the EDB, provision for user-friendly software to access and process the data, exploration and/or confirmation of embrittlement prediction models, provision for rapid investigation of regulatory issues, and provision for the technical bases for voluntary consensus standards or regulatory guides. The EDB is designed for use with a personal computer. The data are collected into "raw data files." Traceability of all data is maintained by including complete references along with the page numbers. External data verification of the PR-EDB is the responsibility of the vendors, who were responsible for the insertion and testing of the materials in the surveillance capsules. Internal verification is accomplished by checking against references and checking for inconsistencies. Examples of information contained in the EDBs are: Charpy data, tensile data, reactor type, irradiation environments, fracture toughness data, instrumented Charpy data, pressure-temperature (P-T) data, chemistry data, and material history. The TR-EDB additionally has annealing Charpy data. The current version of the PR-EDB contains the test results from 269 Charpy capsules irradiated in 101 reactors. These results include 320 plate data points, 123 forging data points, 113 standard reference materials (SRMS) or correlation monitor (CM) points, 244 weld material data points, and 220 heat-affected-zone (HAZ) material data points. Similarly, the TR-EDB contains information for 290 SRM or CM points, 342 plate data points, 165 forging data points, 378 welds, and 55 HAZ materials. [copied from http://rsicc.ornl.gov/RelatedLinks.aspx?t=edb

Fission and Fusion energy is matched by the need to completely separate civilian energy programmes from the production of nuclear weapons. The Nuclear Proliferation Treaty (NPT, 1968) muddles these issues together. The case is presented here for making a new Nuclear Energy Security Treaty (NESST) which is rigorous, enforceable without violence, and separate from the political quagmire of nuclear weapons.

A safety assessment code, DecDose, for decommissioning of nuclear facilities has been developed, based on the experiences of the decommissioning project of Japan Power Demonstration Reactor (JPDR) at Japan Atomic Energy Research Institute (currently JAEA). DecDose evaluates the annual exposure dose of the public and workers according to the progress of decommissioning, and also evaluates the public dose at accidental situations including fire and explosion. As for the public, both the internal and the external doses are calculated by considering inhalation, ingestion, direct radiation from radioactive aerosols and radioactive depositions, and skyshine radiation from waste containers. For external dose for workers, the dose rate from contaminated components and structures to be dismantled is calculated. Internal dose for workers is calculated by considering dismantling conditions, e.g. cutting speed, cutting length of the components and exhaust velocity. Estimation models for dose rate and staying time were verified by comparison with the actual external dose of workers which were acquired during JPDR decommissioning project. DecDose code is expected to contribute the safety assessment for decommissioning of nuclear facilities.

The lessons learned from the Fukushima Daiichi accident have focused on preventive measures designed to protect nuclear reactors, and crisis management plans. Although there is still no end in sight to the accident that occurred on March 11, 2011, how engineers have handled the aftermath offers new insight into the capacity of organizations to adapt in situations that far exceed the scope of safety standards based on probabilistic risk assessment and on the comprehensive identification of disaster scenarios. Ongoing crises in which conventional resources are lacking, but societal expectations are high, call for "engineering thinking in emergency situations." This is a new concept that emphasizes adaptability and resilience within organizations-such as the ability to create temporary new organizational structures; to quickly switch from a normal state to an innovative mode; and to integrate a social dimension into engineering activities. In the future, nuclearsafety oversight authorities should assess the ability of plant operators to create and implement effective engineering strategies on the fly, and should require that operators demonstrate the capability for resilience in the aftermath of an accident. PMID:25419015

The lessons learned from the Fukushima Daiichi accident have focused on preventive measures designed to protect nuclear reactors, and crisis management plans. Although there is still no end in sight to the accident that occurred on March 11, 2011, how engineers have handled the aftermath offers new insight into the capacity of organizations to adapt in situations that far exceed the scope of safety standards based on probabilistic risk assessment and on the comprehensive identification of disaster scenarios. Ongoing crises in which conventional resources are lacking, but societal expectations are high, call for “engineering thinking in emergency situations.” This is a new concept that emphasizes adaptability and resilience within organizations—such as the ability to create temporary new organizational structures; to quickly switch from a normal state to an innovative mode; and to integrate a social dimension into engineering activities. In the future, nuclearsafety oversight authorities should assess the ability of plant operators to create and implement effective engineering strategies on the fly, and should require that operators demonstrate the capability for resilience in the aftermath of an accident. PMID:25419015

Questions of whether past nuclear regulatory body of Korea is not a proper system to monitor and check the country's nuclear energy policy and utilization have been raised. Moreover, a feeling of insecurity regarding nuclearsafety after the nuclear accident in Japan has spread across the public. This has stimulated a renovation of the nuclearsafety regime in Korea. The NuclearSafety and Security Commission (NSSC) was launched on October 26, 2011 as a regulatory body directly under the President in charge of strengthening independence and nuclearsafety. This was a meaningful event as the NSSC it is a much more independent regulatory system for Korea. However, the NSSC itself does not guarantee an enhanced public acceptance of the nuclear policy and stable use nuclear energy. This study introduces the new NSSC system and its details in terms of organization structure, appropriateness of specialty, budget stability, and management system. (authors)

This Handbook establishes general safety, reliability, and quality assurance (SR&QA) guidelines for use on flight and ground-based projects conducted at the Ames, Langley, and Lewis Research Centers, hereafter identified as the Office of Aeronautics, Exploration and Technology (OAET) Centers. This document is applicable to all projects and operations conducted at these Centers except for those projects covered by more restrictive provisions such as the Space Shuttle, Space Station, and unmanned spacecraft programs. This Handbook is divided into two parts. The first (Chapters 1 and 2) establishes the SR&QA guidelines applicable to the OAET Centers, and the second (Appendices A, B, C, and D) provides examples and definitions for the total SR&QA program. Each center should implement SR&QA programs using these guidelines with tailoring appropriate to the special projects conducted by each Center. This Handbook is issued in loose-leaf form and will be revised by page changes.

This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces.

The NPP is used to planning maintenance, in-service inspection, surveillance test, fuel handling and design modification in the refueling outage; the operator response capability will be reduced plus some of the plant systems out of service or loss of power at this time. Based on 8 times refueling outage experiences of the Qinshan NPP, this article provide some good practice and lesson learned for the nuclearsafety risk management focus at four safety function areas of Residual Heat Removal Capability, Inventory Control, Power availability and Reactivity control. (authors)

Nuclear driven engines are described that could be run in either pulsed or steady state modes. In the pulsed mode nuclear energy is released by fissioning of uranium or plutonium in a supercritical assembly of fuel and working gas. In a steady state mode a fuel-gas mixture is injected into a magnetic nozzle where it is compressed into a critical state and produces energy. Engine performance is modeled using a code that calculates hydrodynamics, fission energy production, and neutron transport self-consistently. Results are given demonstrating a large negative temperature coefficient that produces self-shutoff or control of energy production. Reduced fission product inventory and the self-shutoff provide inherent nuclearsafety. It is expected that nuclear engine reactor units could be scaled up from about 100 MW{sub e}.

This paper illustrates some of the promise and needed future work for risk-informed, performance-based regulation (RIPBR). RIPBR is an evolving alternative to the current prescriptive method of nuclearsafety regulation. Prescriptive regulation effectively constitutes a long, fragmented checklist of requirements that safety-related systems in a plant must satisfy. RIPBR, instead, concentrates upon satisfying negotiated performance goals and incentives for judging and rewarding licensee behavior to improve safety and reduce costs. In a project reported here, a case study was conducted concerning a pressurized water reactor (PWR) emergency diesel generator (EDG). Overall, this work has shown that the methods of RIPBR are feasible to use, and capable of justifying simultaneous safety and economic nuclear power improvements. However, it also reveals several areas where the framework of RIPBR should be strengthened. First, researchers need better data and understanding regarding individual component-failure modes that may cause components to fail. Not only are more data needed on failure rates, but more data and understanding are needed to enable analysts to evaluate whether these failures become more likely as the interval between tests is increased. This is because the current state of failure data is not sufficiently finely detailed to define the failure rates of individual component failure modes; such knowledge is needed when changing component-specific regulatory requirements. Second, the role of component testing, given that a component has failed, needs to be strengthened within the context of RIPBR. This includes formulating requirements for updating the prior probability distribution of a component failure rate and conducting additional or more frequent testing. Finally, as a means of compensating for unavoidable uncertainty as an obstacle to regulatory decision-making, limits to knowledge must be treated explicitly and formally. This treatment includes the

This report evaluates nuclear criticality safety for Spray Booth Operations in the Decontamination and Recovery Facility, X-705, at the Portsmouth Gaseous Diffusion Plant. A general description of current procedures and related hardware/equipment is presented. Control parameters relevant to nuclear criticality safety are explained, and a consolidated listing of administrative controls and safety systems is developed. Based on compliance with DOE Orders and MMES practices, the overall operation is evaluated, and recommendations for enhanced safety are suggested.

... Inventory AGENCY: Defense Nuclear Facilities Safety Board (Board). ACTION: Notice of public availability of... show how contracted resources are distributed throughout the agency. The inventory has been developed... From the Federal Register Online via the Government Publishing Office DEFENSE NUCLEAR...

Covariance data computational methods and data used for sensitivity and uncertainty analysis within the SCALE nuclear analysis code system are presented. Applications in criticality safety calculations and used nuclear fuel analysis are discussed.

Covariance data computational methods and data used for sensitivity and uncertainty analysis within the SCALE nuclear analysis code system are presented. Applications in criticality safety calculations and used nuclear fuel analysis are discussed.

Recent focus on the need to improve the quality and safety of health care has created new challenges for academic health centers (AHCs). Whereas previously quality was largely assumed, today it is increasingly quantifiable and requires organized systems for improvement. Traditional structures and cultures within AHCs, although well suited to the tripartite missions of teaching, research, and clinical care, are not easily adaptable to the tasks of measuring, reporting, and improving quality. Here, the authors use a case study of Massachusetts General Hospital's efforts to restructure quality and safety to illustrate the value of beginning with a focus on organizational culture, using a systematic process of engaging clinical leadership, developing an organizational framework dependent on proven business principles, leveraging focus events, and maintaining executive dedication to execution of the initiative. The case provides a generalizable example for AHCs of how applying explicit management design can foster robust organizational change with relatively modest incremental financial resources. PMID:19940570

This article presents a practical implementation of a safety verification framework for nuclear power plants (NPPs) based on fuzzy logic where hazard scenarios are identified in view of safety and control limits in different plant process values. Risk is estimated quantitatively and compared with safety limits in real time so that safety verification can be achieved. Fuzzy logic is used to define safety rules that map hazard condition with required safety protection in view of risk estimate. Case studies are analyzed from NPP to realize the proposed real-time safety verification framework. An automated system is developed to demonstrate the safety limit for different hazard scenarios. PMID:23020592

The ambulatory health care center has an empirical need to maintain adequate policies and procedures designed to safeguard the health and physical safety of both clients and employees. Of particular concern is the health care worker's risk of injury or exposure to infection. This article discusses the development of a comprehensive employee-safety and infection-control manual. Such a manual can assist in the accreditation of an ambulatory health care center by a national accrediting organization, and can also serve as a risk-management tool. An example of a manual's table of contents, a specific policy/procedure, a waiver form regarding human immunodeficiency virus exposure, and a hepatitis B virus "vaccination status" form are provided. Additionally, two "quick-look" texts are presented as examples of fast reference guides for common safety practices and the treatment of the employee with a health-threatening exposure. PMID:1905791

One of the major outgrowths resulting from the collapse of the former Soviet Union (FSU) has been an increase in technical information exchange and dialogue between the Russian and American nuclear weapons laboratories. One area of such discussions is concerned with the safety of low probability, high consequence systems and operations. In order to further the understanding between the respective institutes in this important area, a collaborative effort has been established between Sandia National Laboratories and the two premier Russian nuclear weapons laboratories, Arzamas-16 and Chelyabinsk-70, in which a common database has been developed which contains safety information provided by all three laboratories. More than 1,200 documents have been placed by the three institutes into this database. This paper describes the details of this data base, including the types of safety information being stored.

The Buffalo Materials Research Center (BMRC) is located on the campus of the State University of New York at Buffalo, Principal facilities within BMRC include a 2-MW PULSTAR, low-enrichment reactor, an electron accelerator, and irradiated materials remote testing facilities. The reactor and the materials testing facilities have been utilized extensively in support of the power reactor community since 1961. This paper briefly highlights the nature and scope of this service. The BMRC is operated for the university by Buffalo Materials Research, Inc., a private for-profit company, which is a subsidiary of Materials Engineering Associates, Inc. (MEA), a Maryland-based materials testing company. A primary mission of MEA has been research on the effects of neutron irradiation on reactor structural materials, including those used for pressure vessel and piping systems. The combined resources of MEA and BMRC have played a pivotal role in the assessment of reactor pressure vessel safety both in the United States and abroad and in the development of new radiation-resistant steels.

This PhD dissertation focuses on the improvement of very high temperature thermometry (1100ºC to 2480ºC), with special emphasis on the application to the field of nuclear reactor safety and severe accident research. Two main projects were undertaken to achieve this objective: -The development, testing and transposition of high-temperature fixed point (HTFP) metal-carbon eutectic cells, from metrology laboratory precision (+/-0.001ºC) to applied research with a reasonable degradation of uncertainties (+/-3-5ºC). -The corrosion study and metallurgical characterization of Type-C thermocouple (service temp. 2300ºC) prospective sheath material was undertaken to extend the survivability of TCs used for molten metallic/oxide corium thermometry (below 2000ºC).

The Rhode Island Nuclear Science Center (RINSC), located on the Narragansett Bay Campus of the University of Rhode Island, is a state-owned and US NRC-licensed nuclear facility constructed for educational and industrial applications. The main building of RINSC houses a two-megawatt (2 MW) thermal power critical reactor immersed in demineralized water within a shielded tank. As its original design in 1958 by the Rhode Island Atomic Energy Commission focused on the teaching and research use of the facility, only a minimum of 3.85 kg fissile uranium-235 was maintained in the fuel elements to allow the reactor to reach a critical state. In 1986 when RINSC was temporarily shutdown to start US DOE-directed core conversion project for national security reasons, all the U-Al based Highly-Enriched Uranium (HEU, 93% uranium-235 in the total uranium) fuel elements were replaced by the newly developed U{sub 3}Si{sub 2}-Al based Low Enriched Uranium (LEU, {le}20% uranium-235 in the total uranium) elements. The reactor first went critical after the core conversion was achieved in 1993, and feasibility study on the core upgrade to accommodate Boron Neutron-Captured Therapy (BNCT) was completed in 2000 [3]. The 2-MW critical reactor at RINSC which includes six beam tubes, a thermal column, a gamma-ray experimental station and two pneumatic tubes has been extensive utilized as neutron-and-photon dual source for nuclear-specific research in areas of material science, fundamental physics, biochemistry, and radiation therapy. After the core conversion along with several major system upgrade (e.g. a new 3-MW cooling tower, a large secondary piping system, a set of digitized power-level instrument), the reactor has become more compact and thus more effective to generate high beam flux in both the in-core and ex-core regions for advance research. If not limited by the manpower and operating budget in recent years, the RINSC built ''in concrete'' structure and control systems should have

The purpose of Volume 1 of this 4-volume style guide is to furnish guidelines on writing and publishing Safety Analysis Reports (SARs) for DOE nuclear facilities at Sandia National Laboratories. The scope of Volume 1 encompasses not only the general guidelines for writing and publishing, but also the prescribed topics/appendices contents along with examples from typical SARs for DOE nuclear facilities.

The purpose of this paper is to recommend regulatory guidance for reviewers examining real-time performance of computer-based safety systems used in nuclear power plants. Three areas of guidance are covered in this report. The first area covers how to determine if, when, and what prototypes should be required of developers to make a convincing demonstration that specific problems have been solved or that performance goals have been met. The second area has recommendations for timing analyses that will prove that the real-time system will meet its safety-imposed deadlines. The third area has description of means for assessing expected or actual real-time performance before, during, and after development is completed. To ensure that the delivered real-time software product meets performance goals, the paper recommends certain types of code-execution and communications scheduling. Technical background is provided in the appendix on methods of timing analysis, scheduling real-time computations, prototyping, real-time software development approaches, modeling and measurement, and real-time operating systems.

This Handbook provides detailed guidance and procedures to implement the General Statement of DOE Enforcement Policy (Enforcement Policy or Policy). A copy of this Enforcement Policy is included for ready reference in Appendix D. The guidance provided in this Handbook is qualified, however, by the admonishment to exercise discretion in determining the proper disposition of each potential enforcement action. As discussed in subsequent chapters, the Enforcement and Investigation Staff will apply a number of factors in assessing each potential enforcement situation. Enforcement sanctions are imposed in accordance with the Enforcement Policy for the purpose of promoting public and worker health and safety in the performance of activities at DOE facilities by DOE contractors (and their subcontractors and suppliers) who are indemnified under the Price-Anderson Amendments Act. These indemnified contractors, and their suppliers and subcontractors, will be referred to in this Handbook collectively as DOE contractors. It should be remembered that the purpose of the Department`s enforcement policy is to improve nuclearsafety for the workers and the public, and this goal should be the prime consideration in exercising enforcement discretion.

This document is the ninth in a series of compilations of fission yield data made at Vallecitos NuclearCenter in which fission yield measurements reported in the open literature and calculated charge distributions have been utilized to produce a recommended set of yields for the known fission products. The original data with reference sources, as well as the recommended yields are presented in tabular form for the fissionable nuclides U-235, Pu-239, Pu-241, and U-233 at thermal neutron energies; for U-235, U-238, Pu-239, and Th-232 at fission spectrum energies; and U-235 and U-238 at 14 MeV. In addition, U-233, U-236, Pu-240, Pu-241, Pu-242, Np-237 at fission spectrum energies; U-233, Pu-239, Th-232 at 14 MeV and Cf-252 spontaneous fission are similarly treated. For 1979 U234F, U237F, Pu249H, U234He, U236He, Pu238F, Am241F, Am243F, Np238F, and Cm242F yields were evaluated. In 1980, Th227T, Th229T, Pa231F, Am241T, Am241H, Am242Mt, Cm245T, Cf249T, Cf251T, and Es254T are also evaluated.

ENSDF contains evaluated nuclear structure and decay data in a standard format. An international network of evaluators contributes to the database, which is maintained by the National Nuclear Data Center at Brookhaven National Laboratory. Information in the database is regularly updated to reflect revised evaluation results. Most of the recently completed evaluations are published in Nuclear Data Sheets, a monthly journal published by Academic Press, a division of Elsevier Science. For each nuclide, all known experimental data used to deduce nuclear structure information are included. Each type of experiment is presented as a separate dataset. In addition, there is a dataset of "adopted" level and gamma-ray transition properties, which represent the evaluator's determination of the best values for these properties, based on all available experimental data. As of February 2008, the ENSDF database contains 16236 datasets for 3030 nuclides. (Taken from the NNDC's information page on ENSDF at http://www.nndc.bnl.gov/ensdf/ensdf_info.jsp) ENSDF may be browsed or the data may be retrieved based on nuclide, charge, or mass, or by indexed reaction and decay quantities. (Specialized interface)

Background: Healthcare professionals encounter many medical risks while providing healthcare services to individuals and the community. Thus, occupational safety studies are very important in health care organizations. They involve studies performed to establish legal, technical, and medical measures that must be taken to prevent employees from sustaining physical or mental damage because of work hazards. Objectives: This study was conducted to determine if the occupational safety of health personnel at community and family health centers (CHC and FHC) has been achieved. Martials and Methods: The population of this cross-sectional study comprised 507 nurses, 199 physicians, and 237 other medical personnel working at a total of 18 family health centers (FHC) and community health centers (CHC) in Trabzon, Turkey. The sample consisted of a total of 418 nurses, 156 physicians, and 123 other medical personnel. Sampling method was not used, and the researchers tried to reach the whole population. Data were gathered with the Occupational Safety Scale (OSS) and a questionnaire regarding demographic characteristics and occupational safety. Results: According to the evaluations of all the medical personnel, the mean ± SD of total score of the OSS was 3.57 ± 0.98; of the OSS’s subscales, the mean ± SD of the health screening and registry systems was 2.76 ± 1.44, of occupational diseases and problems was 3.04 ± 1.3 and critical fields control was 3.12 ± 1.62. In addition, occupational safety was found more insufficient by nurses (F = 14.18; P < 0.001). Conclusions: All healthcare personnel, particularly nurses working in CHCs and FHCs found occupational safety to be insufficient as related to protective and supportive activities. PMID:25558383

The Nuclear Installations Inspectorate (NII) of the UK's Health and Safety Executive (HSE) has completed a review of their Safety Assessment Principles (SAPs) for Nuclear Installations recently. During the period of the SAPs review in 2004-2005 the designers of future UK naval reactor plant were optioneering the control and protection systems that might be implemented. Because there was insufficient regulatory guidance available in the naval sector to support this activity the Defence NuclearSafety Regulator (DNSR) invited the NII to collaborate with the production of a guidance document that provides clarity of regulatory expectations for the production of safety cases for computer based safety systems. A key part of producing regulatory expectations was identifying the relevant extant standards and sector guidance that reflect good practice. The three principal sources of such good practice were: IAEA Safety Guide NS-G-1.1 (Software for Computer Based Systems Important to Safety in Nuclear Power Plants), European Commission consensus document (Common Position of European Nuclear Regulators for the Licensing of Safety Critical Software for Nuclear Reactors) and IEC nuclear sector standards such as IEC60880. A common understanding has been achieved between the NII and DNSR and regulatory guidance developed which will be used by both NII and DNSR in the assessment of computer-based safety systems and in the further development of more detailed joint technical assessment guidance for both regulatory organisations. (authors)

This paper represents a chapter of the International Astronautical Academy's Cosmic Study on safety, legal and policy aspects of advanced (specifically nuclear) power and propulsions systems; it is divided into several sections. The first section covers a series of findings and develops a set of recommendations for operations of space reactor systems in a safe, environmentally compliant fashion. The second section develops a generic set of hazard scenarios that might be experienced by a space nuclear system with emphasis on different methods under which such a system could be engaged, such as surface power, in-space nuclear electric or nuclear thermal propulsion. The third section develops these into test and analysis efforts that would likely be conducted. Risk areas with engineering judgment set toward frequency and consequences. The fourth section identifies what probable technology limits might be experienced by nuclear propulsion systems and the exploration limitations these technology restrictions might impose. Where the IAA recommends a change, the IAA leadership should be prepared to work with national and international bodies to implement the desired modifications.

This document is a review journal that covers significant developments in the field of nuclearsafety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

This document contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE non-reactor nuclear facilities. Adherence to these guidelines will provide consistency and uniformity in criticality safety evaluations (CSEs) across the complex and will document compliance with the requirements of DOE Order 5480.24.

Although unpublicized, the use of quantitative safety goals and probabilistic reliability analysis for licensing nuclear reactors has become a reality in the United Kingdom. This conclusion results from an examination of the process leading to the licensing of the Sizewell B PWR in England. The licensing process for this reactor has substantial implications for nuclearsafety standards in Britain, and is examined in the context of the growing trend towards quantitative safety goals in the United States. PMID:3685540

Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications.

The ITER Gas Injection System (GIS) plays an important role on fueling, wall conditioning and distribution for plasma operation. Besides that, to support the safety function of ITER, GIS needs to implement three nuclearsafety Instrumentation and Control (I&C) functions. In this paper, these three functions are introduced with the emphasis on their latest safety classifications. The nuclear I&C design concept is briefly discussed at the end.

Nuclear energy continues to be a strong and growing component of economic development in Northeast Asia. A broad range of nuclear energy systems already exists across the region and vigorous growth is projected. Associated with these capabilities and plans are various concerns about operational safety, environmental protection, and accumulation of spent fuel and other nuclear materials. We consider cooperative measures that might address these concerns. The confidence building measures suggested here center on the sharing of information to lessen concerns about nuclear activities or to solve technical problems. These activities are encompassed by an Enhanced Nuclear Transparency in Northeast Asia (ENTNEA) concept that would be composed of near-term, information-sharing activities and an eventual regional institution. The near-term activities would address specific concerns and build a tradition of cooperation; examples include radiation measurements for public safety and emergency response, demonstration of safe operations at facilities and in transportation, and material security in the back end of the fuel cycle. Linkages to existing efforts and organizations would be sought to maximize the benefits of cooperation. In the longer term, the new cooperative tradition might evolve into an ENTNEA institution. In institutional form, ENTNEA could combine the near-term activities and new cooperative activities, which might require an institutional basis, for the mutual benefit and security of regional parties.

Objective: Most of the radiopharmaceuticals used in nuclear medicine are excreted via the urinary system. This study evaluated the importance of a reduction in bladder radioactivity for radiation safety. Methods: The study group of 135 patients underwent several organ scintigraphies [40/135; thyroid scintigraphy (TS), 30/135; whole body bone scintigraphy (WBS), 35/135; myocardial perfusion scintigraphy (MPS) and 30/135; renal scintigraphy (RS)] by a technologist within 1 month. In full and empty conditions, static bladder images and external dose rate measurements at 0.25, 0.50, 1, 1.5 and 2 m distances were obtained and decline ratios were calculated from these two data sets. Results: External radiation dose rates were highest in patients undergoing MPS. External dose rates at 0.25 m distance for TS, TKS, MPS and BS were measured to be 56, 106, 191 and 72 μSv h-1 for full bladder and 29, 55, 103 and 37 μSv h-1 for empty bladder, respectively. For TS, WBS, MPS and RS, respectively, average decline ratios were calculated to be 52%, 55%, 53% and 54% in the scintigraphic assessment and 49%, 51%, 49%, 50% and 50% in the assessment with Geiger counter. Conclusion: Decline in bladder radioactivity is important in terms of radiation safety. Patients should be encouraged for micturition after each scintigraphic test. Spending time together with radioactive patients at distances less than 1 m should be kept to a minimum where possible. Conflict of interest:None declared. PMID:24416625

The NASA Engineering and SafetyCenter (NESC) funded a proposal to achieve initial basic acoustic characterization of ML (melamine) foam, which could serve as a starting point for a future, more comprehensive acoustic test program for ML foam. A project plan was developed and implemented to obtain acoustic test data for both normal and enhanced ML foam. This project became known as the NESC Enhanced Melamine Foam Acoustic Test (NEMFAT). This document contains the outcome of the NEMFAT project.

This paper describes a study, which had as its principal goal the development of a sufficiently detailed 5 x 5 Risk Matrix Scorecard. The purpose of this scorecard is to outline the criteria by which technical issues can be qualitatively and initially prioritized. The tool using this score card has been proposed to be one of the information resources the NASA Engineering and SafetyCenter (NESC) takes into consideration when making decisions with respect to incoming information on safety concerns across the entire NASA agency. The contents of this paper discuss in detail each element of the risk matrix scorecard, definitions for those elements and the rationale behind the development of those definitions. This scorecard development was performed in parallel with the tailoring of the existing Futron Corporation Integrated Risk Management Application (IRMA) software tool. IRMA was tailored to fit NESC needs for evaluating incoming safety concerns and was renamed NESC Assessment Risk Management Application (NAFMA) which is still in developmental phase.

With the increasing interests in the nuclear energy, new nuclear power plants will be constructed and licensed, and older generation ones will be upgraded for assuring continuing operation. The tendency of adopting the latest proven technology and the fact of older parts becoming obsolete have made the upgrades imperative. One of the areas for upgrades is the older CRT display being replaced by the latest graphics displays running under modern real time operating system (RTOS) with safety graded modern computer. HFC has developed a graphic display utility (GDU) under the QNX RTOS. A standard off-the-shelf software with a long history of performance in industrial applications, QNX RTOS used for safety applications has been examined via a commercial dedication process that is consistent with the regulatory guidelines. Through a commercial survey, a design life cycle and an operating history evaluation, and necessary tests dictated by the dedication plan, it is reasonably confirmed that the QNX RTOS was essentially equivalent to what would be expected in the nuclear industry. The developed GDU operates and communicates with the existing equipment through a dedicated serial channel of a flat panel controller (FPC) module. The FPC module drives a flat panel display (FPD) monitor. A touch screen mounted on the FPD serves as the normal operator interface with the FPC/FPD monitor system. The GDU can be used not only for replacing older CRTs but also in new applications. The replacement of the older CRT does not disturb the function of the existing equipment. It not only provides modern proven technology upgrade but also improves human ergonomics. The FPC, which can be used as a standalone controller running with the GDU, is an integrated hardware and software module. It operates as a single board computer within a control system, and applies primarily to the graphics display, targeting, keyboard and mouse. During normal system operation, the GDU has two sources of data

Extreme storm events over the last 35 years are being assessed to evaluate flood estimates for safety assessments of dams, nuclear power plants, and other high-hazard structures in the U.S. The current storm rainfall design standard for evaluating the flood potential at dams and non-coastal nuclear power plants is the Probable Maximum Precipitation (PMP). PMP methods and estimates are published in the National Weather Service generalized hydrometeorological reports (HMRs). A new Federal Interagency cooperative effort is reviewing hydrometeorologic data from large storms which have occurred in the last 20 to 40 years and were not included in the database used in the development of many of the HMRs. Extreme storm data, such as the January 1996 storm in Pennsylvania, June 2008 Iowa storms, and Hurricanes Andrew (1992), Floyd (1999), Isabel (2003), Katrina (2005), need to be systematically assembled and analyzed for use in these regional extreme storm studies. Storm maximization, transposition, envelopment, and depth-area duration procedures will incorporate recent advances in hydrometeorology, including radar precipitation data and stochastic storm techniques. We describe new cooperative efforts to develop a database of extreme storms and to examine the potential impacts of recent extreme storms on PMP estimates. These efforts will be coordinated with Federal agencies, universities, and the private sector through an Extreme Storm Events Work Group under the Federal Subcommittee on Hydrology. This work group is chartered to coordinate studies and develop databases for reviewing and improving methodologies and data collection techniques used to estimate design precipitation up to and including the PMP. The initial effort focuses on collecting and reviewing extreme storm event data in the Southeastern U.S. that have occurred since Tropical Storm Agnes (1972). Uncertainties and exceedance probability estimates of PMP are being explored. Potential effects of climate

Depletion calculations have been performed for the McClellan reactor history from January 1990 through August 1996. A database has been generated for continuing use by operations personnel which contains the isotopic inventory for all fuel elements and fuel-followed control rods maintained at McClellan. The calculations are based on the three-dimensional diffusion theory code REBUS-3 which is available through the Radiation Safety Information Computational Center (RSICC). Burnup-dependent cross-sections were developed at zero power temperatures and full power temperatures using the WIMS code (also available through RSICC). WIMS is based on discretized transport theory to calculate the neutron flux as a function of energy and position in a one-dimensional cell. Based on the initial depletion calculations, a method was developed to allow operations personnel to perform depletion calculations and update the database with a minimal amount of effort. Depletion estimates and calculations can be performed by simply entering the core loading configuration, the position of the control rods at the start and end of cycle, the reactor power level, the duration of the reactor cycle, and the time since the last reactor cycle. The depletion and buildup of isotopes of interest (heavy metal isotopes, erbium isotopes, and fission product poisons) are calculated for all fuel elements and fuel-followed control rods in the MNRC inventory. The reactivity loss from burnup and buildup of fission product poisons and the peak xenon buildup after shutdown are also calculated. The reactivity loss from going from cold zero power to hot full power can also be calculated by using the temperature-dependent, burnup-dependent cross-sections. By calculating all of these reactivity effects, operations personnel are able to estimate the total excess reactivity necessary to run the reactor for the given cycle. This method has also been used to estimate the worth of individual control rods. Using this

The research effort is a part of the Light Water Reactor Sustainability (LWRS) Program. LWRS is a research and development program sponsored by the Department of Energy, performed in close collaboration with industry to provide the technical foundations for licensing and managing the long-term, safe and economical operation of current nuclear power plants. The LWRS Program serves to help the US nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. The Outage Control Center (OCC) Pilot Project was directed at carrying out the applied research for development and pilot of technology designed to enhance safe outage and maintenance operations, improve human performance and reliability, increase overall operational efficiency, and improve plant status control. Plant outage management is a high priority concern for the nuclear industry from cost and safety perspectives. Unfortunately, many of the underlying technologies supporting outage control are the same as those used in the 1980’s. They depend heavily upon large teams of staff, multiple work and coordination locations, and manual administrative actions that require large amounts of paper. Previous work in human reliability analysis suggests that many repetitive tasks, including paper work tasks, may have a failure rate of 1.0E-3 or higher (Gertman, 1996). With between 10,000 and 45,000 subtasks being performed during an outage (Gomes, 1996), the opportunity for human error of some consequence is a realistic concern. Although a number of factors exist that can make these errors recoverable, reducing and effectively coordinating the sheer number of tasks to be performed, particularly those that are error prone, has the potential to enhance outage efficiency and safety. Additionally, outage management requires precise coordination of work groups that do not always share similar objectives. Outage

The Nuclear Weapons Reduction Initiative announced by President Bush on September 27, 1991, is described herein as set forth in Defense Secretary Cheney`s Nuclear Arsenal Reduction Order issued September 28, 1991. The implications of the Bush Initiative for improved nuclear weapons safety are assessed in response to a request by US Senators Harkin, Kennedy, and Wirth to the Lawrence Livermore National Laboratory that the author prepare such an assessment. The author provides an estimate of the number of nuclear tests needed to accomplish a variety of specified warhead safety upgrades, then uses the results of this estimate to answer three questions posed by the Senators. These questions concern pit reuse and the number of nuclear tests needed for specified safety upgrades of those ballistic missiles not scheduled for retirement, namely the Minuteman III, C4, and D5 missiles.

Sandia National Laboratories (SNL) and the Nuclear Security Science and Policy Institute (NSSPI) at Texas A&M University are working with Middle East regional partners to set up a nuclear energy safety, safeguards, and security educational institute in the Gulf region. SNL and NSSPI, partnered with the Khalifa University of Science, Technology, and Research (KUSTAR), with suppot from its key nuclear stakeholders, the Emirates Nuclear Energy Corporation (ENEC), and the Federal Authority for Nuclear Regulation (FANR), plan to jointly establish the institute in Abu Dhabi. The Gulf Nuclear Energy Infrastructure Institute (GNEII) will be a KUSTAR-associated, credit-granting regional education program providing both classroom instruction and hands-on experience. The ultimate objective is for GNEII to be autonomous - regionally funded and staffed with personnel capable of teaching all GNEII courses five years after its inauguration. This is a strategic effort to indigenize a responsible nuclear energy culture - a culture shaped by an integrated understanding of nuclearsafety, safeguards and security - in regional nuclear energy programs. GNEII also promotes international interests in developing a nuclear energy security and safety culture, increases collaboration between the nuclear energy security and safety communities, and helps to enhance global standards for nuclear energy technology in the Middle East.

Sandia National Laboratories (SNL) and the Nuclear Security Science and Policy Institute (NSSPI) at Texas A&M University are working with Middle East regional partners to set up a nuclear energy safety, safeguards, and security educational institute in the Gulf region. SNL and NSSPI, partnered with the Khalifa University of Science, Technology, and Research (KUSTAR), with suppot from its key nuclear stakeholders, the Emirates Nuclear Energy Corporation (ENEC), and the Federal Authority for Nuclear Regulation (FANR), plan to jointly establish the institute in Abu Dhabi. The Gulf Nuclear Energy Infrastructure Institute (GNEII) will be a KUSTAR-associated, credit-granting regional education program providing both classroom instruction and hands-on experience. The ultimate objective is for GNEII to be autonomous - regionally funded and staffed with personnel capable of teaching all GNEII courses five years after its inauguration. This is a strategic effort to indigenize a responsible nuclear energy culture - a culture shaped by an integrated understanding of nuclearsafety, safeguards and security - in regional nuclear energy programs. GNEII also promotes international interests in developing a nuclear energy security and safety culture, increases collaboration between the nuclear energy security and safety communities, and helps to enhance global standards for nuclear energy technology in the Middle East.

The construction of the Oak Ridge Gaseous Diffusion Plant (now known as the K-25 Site) began during World War H and eventually consisted of five major process buildings: K-25, K-27, K-29, K-31, and K-33. The plant took natural (0.711% {sup 231}U) uranium as feed and processed it into both low-enriched uranium (LEU) and high-enriched uranium (HEU) with concentrations up to {approximately}93% {sup 231}U. The K-25 and K-27 buildings were shut down in 1964, but the rest of the plant produced LEU until 1985. During operation, inleakage of humid air into process piping and equipment caused reactions with gaseous uranium hexafluoride (UF{sub 6}) that produced nonvolatile uranyl fluoride (UO{sub 2}F{sub 2}) deposits. As part of shutdown, most of the uranium was evacuated as volatile UF{sub 6}. The UO{sub 2}F{sub 2} deposits remained. The U.S. Department of Energy has mitiated a program to unprove nuclear criticality safety by removing the larger enriched uranium deposits.

This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64. PMID:22021060

The key elements of the optimization practice as applied to radiation protection are the monetary value of the averted person-sievert and the aversion coefficient. Determination of the monetary value of the unit averted person-sievert (as alpha(base)-parameter) in Hungary was presented in a previous paper. The estimation of this parameter was carried out by the willingness-to-pay (WTP) method associated with averted occupational exposure (at the NPP Paks/Hungary). The aversion coefficient predicts the importance of dose reduction based on the magnitude of the dose. The assessment of the aversion coefficient occurred also by means of the WTP method in the spring of 2000. Its value has been estimated on the basis of individual preferences concerning the distribution of individual exposure in nuclearsafety. The results achieved by the WTP among the radiation specialists from the NPP Paks, Hungary, assessed a value for the aversion coefficient of 1.86 over the whole range of individual exposure levels. This value is a bit greater than the value obtained in France (1.7) and the higher coefficient expresses a higher priority to reduce the highest individual exposures. PMID:12046754

The Jayne Haines Center for Pharmacogenomics and Drug Safety operates in the Temple University School of Pharmacy and serves as an educational and research facility for professional pharmacy students, graduate students, residents, postdoctoral fellows and faculties. The Center is involved in educational and investigational projects in a setting that includes an inner city research/teaching hospital and the Temple University Schools of Pharmacy, Medicine, Dentistry and Health Professions. The mission of the Haines Center is to facilitate the basic science approach to the problems of pharmacotherapy, and to provide education for future healthcare professionals in the genetics of drug response. With the expanding role of genetic analysis in preventing adverse effects of pharmacotherapy, we are working towards truly personalized pharmacotherapy that fully exploits the advances of modern biomedical science. PMID:23556444

In May 1973 the University of New Mexico conducted the first nationwide criticality safety training and education week-long short course for nuclear criticality safety engineers. Subsequent to that course, the Los Alamos Critical Experiments Facility (LACEF) developed very successful 'hands-on' subcritical and critical training programs for operators, supervisors, and engineering staff. Since the inception of the US Department of Energy (DOE) Nuclear Criticality Technology and Safety Project (NCT&SP) in 1983, the DOE has stimulated contractor facilities and laboratories to collaborate in the furthering of nuclear criticality as a discipline. That effort included the education and training of nuclear criticality safety engineers (NCSEs). In 1985 a textbook was written that established a path toward formalizing education and training for NCSEs. Though the NCT&SP went through a brief hiatus from 1990 to 1992, other DOE-supported programs were evolving to the benefit of NCSE training and education. In 1993 the DOE established a Nuclear Criticality Safety Program (NCSP) and undertook a comprehensive development effort to expand the extant LACEF 'hands-on' course specifically for the education and training of NCSEs. That successful education and training was interrupted in 2006 for the closing of the LACEF and the accompanying movement of materials and critical experiment machines to the Nevada Test Site. Prior to that closing, the Lawrence Livermore National Laboratory (LLNL) was commissioned by the US DOE NCSP to establish an independent hands-on NCSE subcritical education and training course. The course provided an interim transition for the establishment of a reinvigorated and expanded two-week NCSE education and training program in 2011. The 2011 piloted two-week course was coordinated by the Oak Ridge National Laboratory (ORNL) and jointly conducted by the Los Alamos National Laboratory (LANL) classroom education and facility training, the Sandia National

'Defence in Depth' approach constitutes the basis of the IAEA safety standards for nuclear power plants. Lessons learned from the current generation of reactors suggest that, for the next generation of reactor designs, the Defence in Depth philosophy should be retained, and that its implementation should be guided by the probabilistic insights. Recent developments in the area of general safety requirements based on Defence in Depth approach are examined and summarized. Global efforts to harmonize safety requirements for evolutionary nuclear power plants have involved many countries and organizations such as IAEA, US EPRI and European Utility EUR Organization. In recent years, developments of innovative nuclear power plants are also being discussed. The IAEA is currently developing a safety approach specifically for innovative nuclear reactors. This approach will eventually lead to a proposal of safety requirements for innovative reactors. Such activities related to safety requirements of evolutionary and innovative reactors are introduced. Various evolutionary and innovative reactor designs are reported in the world. The safety design features of evolutionary large LWRs, innovative LWRs, Modular High Temperature Gas Reactors and Small Liquid Metal Cooled LMRs are also introduced. Enhanced safety features proposed in such reactors are discussed and summarized according to the levels of Defence in Depth. For future nuclear plants, international cooperation and harmonization, especially in the area of safety, appear to be inevitable. Based on the past experience with many member states, the IAEA believes itself to be the uniquely positioned international organization to play this key role. (authors)

A dry storage array of used nuclear fuel at the Idaho National Laboratory contains a mixture of more than twenty different research and test reactor fuel types in up to 636 fuel storage canisters. New analysis demonstrates that the current arrangement of the different fuel-type canisters does not minimize the system neutron multiplication factor (keff), and that the entire facility storage capacity cannot be utilized without exceeding the subcritical limit (ksafe) for ensuring nuclear criticality safety. This work determines a more optimal arrangement of the stored fuels with a goal to minimize the system keff, but with a minimum of potential fuel canister relocation movements. The solution to this multiple-objective optimization problem will allow for both an improvement in the facility utilization while also offering an enhancement in the safety margin. The solution method applies stochastic approximation and a Tabu search metaheuristic to an empirical model developed from supporting MCNP calculations. The results establish an optimal relocation of between four to sixty canisters, which will allow the current thirty-one empty canisters to be used for storage while reducing the array keff by up to 0.018 +/- 0.003 relative to the current arrangement.

In the separation and purification of nuclear material, nuclear criticality safety (NCS) is of primary concern. The primary nuclear criticality safety controls utilized by the Savannah River Site (SRS) Separations Facilities involve administrative and process equipment controls. Additional assurance of NCS is obtained by identifying key process hold points where sampling is used to independently verify the effectiveness of production control. Nuclearsafety measurements of samples from these key process locations provide a high degree of assurance that processing conditions are within administrative and procedural nuclearsafety controls. An enhanced procedure management system aimed at making improvements in the quality, safety, and conduct of operation was implemented for NuclearSafety Sample (NSS) receipt, analysis, and reporting. All procedures with nuclearsafety implications were reviewed for accuracy and adequate detail to perform the analytical measurements safely, efficiently, and with the utmost quality. Laboratory personnel worked in a Deliberate Operating'' mode (a systematic process requiring continuous expert oversight during all phases of training, testing, and implementation) to initiate the upgrades. Thus, the effort to revise and review nuclearsafety sample procedures involved a team comprised of a supervisor, chemist, and two technicians for each procedure. Each NSS procedure was upgraded to a Use Every Time'' (UET) procedure with sign-off steps to ensure compliance with each step for every nuclearsafety sample analyzed. The upgrade program met and exceeded both the long and short term customer needs by improving measurement reliability, providing objective evidence of rigid adherence to program principles and requirements, and enhancing the system for independent verification of representative sampling from designated NCS points.

In the separation and purification of nuclear material, nuclear criticality safety (NCS) is of primary concern. The primary nuclear criticality safety controls utilized by the Savannah River Site (SRS) Separations Facilities involve administrative and process equipment controls. Additional assurance of NCS is obtained by identifying key process hold points where sampling is used to independently verify the effectiveness of production control. Nuclearsafety measurements of samples from these key process locations provide a high degree of assurance that processing conditions are within administrative and procedural nuclearsafety controls. An enhanced procedure management system aimed at making improvements in the quality, safety, and conduct of operation was implemented for NuclearSafety Sample (NSS) receipt, analysis, and reporting. All procedures with nuclearsafety implications were reviewed for accuracy and adequate detail to perform the analytical measurements safely, efficiently, and with the utmost quality. Laboratory personnel worked in a ``Deliberate Operating`` mode (a systematic process requiring continuous expert oversight during all phases of training, testing, and implementation) to initiate the upgrades. Thus, the effort to revise and review nuclearsafety sample procedures involved a team comprised of a supervisor, chemist, and two technicians for each procedure. Each NSS procedure was upgraded to a ``Use Every Time`` (UET) procedure with sign-off steps to ensure compliance with each step for every nuclearsafety sample analyzed. The upgrade program met and exceeded both the long and short term customer needs by improving measurement reliability, providing objective evidence of rigid adherence to program principles and requirements, and enhancing the system for independent verification of representative sampling from designated NCS points.

The tools and procedures for analyzing environmental quality and safety are reviewed. The process of preparing an environmental impact statement is outlined and the data sources for a safety analysis are discussed. The environmental safety analysis process is demonstrated, using examples from the Galileo, Ulysses, and Venus-earth-earth-gravity-assist programs.

This report documents calculations performed to establish a technical basis for the nuclear criticality safety of favorable geometry containers, sometimes referred to as 5-inch containers, in use at the Portsmouth Gaseous Diffusion Plant. A list of containers currently used in the plant is shown in Table 1.0-1. These containers are currently used throughout the plant with no mass limits. The use of containers with geometries or material types other than those addressed in this evaluation must be bounded by this analysis or have an additional analysis performed. The following five basic container geometries were modeled and bound all container geometries in Table 1.0-1: (1) 4.32-inch-diameter by 50-inch-high polyethylene bottle; (2) 5.0-inch-diameter by 24-inch-high polyethylene bottle; (3) 5.25-inch-diameter by 24-inch-high steel can ({open_quotes}F-can{close_quotes}); (4) 5.25-inch-diameter by 15-inch-high steel can ({open_quotes}Z-can{close_quotes}); and (5) 5.0-inch-diameter by 9-inch-high polybottle ({open_quotes}CO-4{close_quotes}). Each container type is evaluated using five basic reflection and interaction models that include single containers and multiple containers in normal and in credible abnormal conditions. The uranium materials evaluated are UO{sub 2}F{sub 2}+H{sub 2}O and UF{sub 4}+oil materials at 100% and 10% enrichments and U{sub 3}O{sub 8}, and H{sub 2}O at 100% enrichment. The design basis safe criticality limit for the Portsmouth facility is k{sub eff} + 2{sigma} < 0.95. The KENO study results may be used as the basis for evaluating general use of these containers in the plant.

The US Department of Energy (DOE) implements the US Government`s International NuclearSafety Program to improve the level of safety at Soviet-designed nuclear power plants in Central and Eastern Europe, Russia, and Unkraine. The program is conducted consistent with guidance and policies established by the US Department of State (DOS) and the Agency for International Development and in close collaboration with the Nuclear Regulatory Commission. Some of the program elements were initiated in 1990 under a bilateral agreement with the former Soviet Union; however, most activities began after the Lisbon NuclearSafety Initiative was announced by the DOS in 1992. Within DOE, the program is managed by the International Division of the Office of Nuclear Energy. The overall objective of the International NuclearSafety Program is to make comprehensive improvements in the physical conditions of the power plants, plant operations, infrastructures, and safety cultures of countries operating Soviet-designed reactors. This status report summarizes the Internatioal NuclearSafety Program`s activities that have been completed as of September 1994 and discusses those activities currently in progress.

The NASA Engineering and SafetyCenter (NESC), set up in the wake of the Columbia accident to backstop engineers in the space shuttle program, is reviewing hundreds of recurring anomalies that the program had determined don't affect flight safety to see if in fact they might. The NESC is expanding its support to other programs across the agency, as well. The effort, which will later extend to the International Space Station (ISS), is a principal part of the attempt to overcome the normalization of deviance--a situation in which organizations proceeded as if nothing was wrong in the face of evidence that something was wrong--cited by sociologist Diane Vaughn as contributing to both space shuttle disasters.

The U.S. Department of Energy Idaho Operations Office (DOE-ID) provided guidance per DOE-ID Orders 420.C, "Safety Basis Review and Approval Process," and 420.D, "Requirements and Guidance for Safety Analysis," for conducting safety analysis for facilities and activities that do not meet either the nuclear facility criteria or the criteria for not requiring additional safety analysis (NRASA). These facilities and activities are thus designated as "other than nuclear" (OTN), and hazard analyses are performed using a graded approach. This graded approach is done in accordance with DOE-ID Order 420.D. DOE-ID guidance is used to format these OTN facilities and activities into 3-chapter documents, rather than the 17-chapter format specified in DOE-STD-3009-94, "Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports."

The FAR Manual is a convenient easy-to-use collection of the functions, assignments, and responsibilities (FARs) of DOE nuclearsafety personnel. Current DOE directives, including Orders, Secretary of Energy Notices, and other assorted policy memoranda, are the source of this information and form the basis of the FAR Manual. Today, the majority of FARs for DOE personnel are contained in DOE`s nuclearsafety Orders. As these Orders are converted to rules in the Code of Federal Regulations, the FAR Manual will become the sole source for information relating to the functions, assignments, responsibilities of DOE nuclearsafety personnel. The FAR Manual identifies DOE directives that relate to nuclearsafety and the specific DOE personnel who are responsible for implementing them. The manual includes only FARs that have been extracted from active directives that have been approved in accordance with the procedures contained in DOE Order 1321.1B.

Lifting mechanisms safety requirements are discussed in accordance with the present state of development of science and engineering for the protection of life, health, and assets against the dangers of nuclear energy and the ill effects of ionizing radiation.

Current industry and NRC guidance documents such as IEEE 7-4.3.2, Reg. Guide 1.152, and IEEE 603 do not sufficiently define a level of detail for evaluating interdivisional communications independence. The NRC seeks to establish criteria for safety systems communications that can be uniformly applied in evaluation of a variety of safety system designs. This report focuses strictly on communication issues related to data sent between safety systems and between safety and nonsafety systems. Further, the report does not provide design guidance for communication systems nor present detailed failure modes and effects analysis (FMEA) results for existing designs. This letter report describes communications between safety and nonsafety systems in nuclear power plants outside the United States. A limited study of international nuclear power plants was conducted to ascertain important communication implementations that might have bearing on systems proposed for licensing in the United States. This report provides that following information: 1.communications types and structures used in a representative set of international nuclear power reactors, and 2.communications issues derived from standards and other source documents relevant to safety and nonsafety communications. Topics that are discussed include the following: communication among redundant safety divisions, communications between safety divisions and nonsafety systems, control of safety equipment from a nonsafety workstation, and connection of nonsafety programming, maintenance, and test equipment to redundant safety divisions during operation. Information for this report was obtained through publicly available sources such as published papers and presentations. No proprietary information is represented.

The Safety Technology Group is developing methodology that can be used to assess the risk of operating a plant to reprocess spent nuclear fuel. As an early step in the methodology, a preliminary hazards analysis identifies safety-related incidents. In the absence of appropriate safety features, these incidents could lead to significant consequences and risk to onsite personnel or to the public. This report is a compilation of potential safety-related incidents that have been identified in studies at SRL and in safety analyses of various commercially designed reprocessing plants. It is an expanded revision of the version originally published as DP-1558, Published December 1980.

The unique biochemical identity of the nuclear envelope confers its capacity to establish a barrier that protects the nuclear compartment and directly contributes to nuclear function. Recent work uncovered quality control mechanisms employing the endosomal sorting complexes required for transport (ESCRT) machinery and a new arm of endoplasmic reticulum-associated protein degradation (ERAD) to counteract the unfolding, damage, or misassembly of nuclear envelope proteins and ensure the integrity of the nuclear envelope membranes. Moreover, cells have the capacity to recognize and triage defective nuclear pore complexes to prevent their inheritance and preserve the longevity of progeny. These mechanisms serve to highlight the diverse strategies used by cells to maintain nuclear compartmentalization; we suggest they mitigate the progression and severity of diseases associated with nuclear envelope malfunction such as the laminopathies. PMID:26437591

Nitrogen-vacancy (NV) centers in diamond have emerged as a versatile atom-like system, finding diverse applications in metrology and quantum information science, but interaction between the NV center's electronic spin and its nuclear spin environment represent a major source of decoherence. We use optical techniques to monitor and control the nuclear bath surrounding an NV center. Specifically, we create an optical Λ-system using the | +/- 1 > components of the NV center's spin-triplet ground state. When the Zeeman splitting between the two states is equal to the two-photon detuning between the lasers, population is trapped in the resulting dark state. Measuring the rate at which the NV center escapes from the dark state therefore gives information on how spin bath dynamics change the effective magnetic field experienced by the NV center. By monitoring statistics of the emitted photons, we plan to probe non-equilibrium dynamics of the bath.

This report provides a quantitative inventory and composition (chemical and radioactivity) of calcined waste stored at the Idaho Nuclear Technology and Engineering Center. From December 1963 through May 2000, liquid radioactive wastes generated by spent nuclear fuel reprocessing were converted into a solid, granular form called calcine. This report also contains a description of the calcine storage bins.

...The Coast Guard proposes establishing a safety zone on the navigable waters of the Elizabeth River in Norfolk, Portsmouth, and Chesapeake, VA. This action is necessary to provide for the safety of life on navigable waters during the Gilmerton Bridge Center Span Float- in and bridge construction of span placement. This action is intended to restrict vessel traffic movement to protect mariners......

The CEA Cadarache research center is one of the 10 research centers of the French Alternative Energies and Atomic Energy Commission (CEA). Distributed throughout various research platforms, it focuses on nuclear fission, nuclear fusion, new energy technologies (hydrogen, solar, biomass) and fundamental research in the field of vegetal biology. It is the most important technological research and development centers for energy in Europe. Considering the sensitive nature of nuclear activities, the questions surrounding the issue of radioactive waste, the nuclear energy and the social, economic and environmental concerns for present and future generations, the French Government asked nuclear actors to open communication and to give all the information asked by the Local Information Commission (CLI) and the public [1]. In this context, the CEA Cadarache has decided to better show and explain its expertise and experience in the area of nuclear energy and nuclear power plant design, and to make it available to stakeholders and to the public. CEA Cadarache receives each year more than 9000 visitors. To complete technical visits of the research facilities and laboratories, a scientific cultural center has been built in 2011 to inform the public on CEA Cadarache research activities and to facilitate the acceptance of nuclear energy in a way suited to the level of knowledge of the visitors. A modern interactive exhibition of 150 m{sup 2} allows visitors to find out more about energy, CEA Cadarache research programs, radioactive waste management and radiological impact on the research center activities. It also offers an auditorium for group discussions and for school groups to discover science through enjoyment. This communication center has received several thousand visitors since its opening on October 2011; the initial results of this experience are now available. It's possible to explain the design of this exhibition, to give some statistics on the number of the visitors

Passive/inherent safety implies a technical revolution in our approach to nuclear power safety. This direction is discussed herein for light-water reactors (LWRs) -- the predominant type of power reactor used in the world today. At Oak Ridge National Laboratory (ORNL) the approach to the development of passive/inherent safety for LWRs consists of four steps: identify and quantify safety requirements and goals; identify and quantify the technical functional requirements needed for safety; identify, invent, develop, and quantify technical options that meet both of the above requirements; and integrate safety systems into designs of economic and reliable nuclear power plants. Significant progress has been achieved in the first three steps of this program. The last step involves primarily the reactor vendors. These activities, as well as related activities worldwide, are described here. 27 refs., 7 tabs.

The Chairman of the Senate Committee on Govermental Affairs asked the United States General Accounting Office (GAO) to review the safeguards and nuclear power plant safety programs of the International Atomic Energy Agency (IAEA). This report examines (1) the effectiveness of IAEA`s safeguards program and the adequacy of program funding, (2) the management of U.S. technical assistance to the IAEA`s safeguards program, and (3) the effectiveness of IAEA`s program for advising United Nations (UN) member states about nuclear power plant safety and the adequacy of program funding. Under its statute and the Treaty on the Non-Proliferation of Nuclear Weapons, IAEA is mandated to administer safeguards to detect diversions of significant quantities of nuclear material from peaceful uses. Because of limits on budget growth and unpaid contributions, IAEA has had difficulty funding the safeguards program. IAEA also conducts inspections of facilities or locations containing declared nuclear material, and manages a program for reviewing the operational safety of designated nuclear power plants. The U.S. technical assistance program for IAEA safeguards, overseen by an interagency coordinating committee, has enhanced the agency`s inspection capabilities, however, some weaknesses still exist. Despite financial limitations, IAEA is meeting its basic safety advisory responsibilities for advising UN member states on nuclearsafety and providing requested safety services. However, IAEA`s program for reviewing the operational safety of nuclear power plants has not been fully effective because the program is voluntary and UN member states have not requested IAEA`s review of all nuclear reactors with serious problems. GAO believes that IAEA should have more discretion in selecting reactors for review.

This paper shows a comparative analysis between France and Japan on the way of the local governments' involvement in nuclearsafety governance through some interviews. In France, a law came into force that requires related local governments to establish "Commision Locale d'Information" (CLI), which means the local governments officially involve in nuclear regulatory activity. Meanwhile, in Japan, related local governments substantially involve in the operation of nuclear facilities through the "safety agreements" in spite of the lack of legal authority. As a result of comparative analysis, we can point out some institutional input from French cases as follows: to clarify the local governments' roles in the nuclear regulation system, to establish the official channels of communication among nuclear utilities, national regulatory authorities and local governments, and to stipulate explicitly the transparency as a purpose of safety regulation.

Transient electron paramagnetic resonance is used to study the secondary radical pair in plant photosystem I. Nuclear coherences are observed in the transverse magnetization at lower temperatures following light excitation. Comparative studies of deuterated and deuterated {sup 15}N-substituted cyanobacteria S. lividus indicate assignment of these coherences to nitrogen nuclei in the primary donor and deuterons in the secondary acceptor. The modulation amplitude of a deuteron matrix line, as a function of the microwave power, reveals a distinct resonance behavior. The maximum amplitude is obtained when the Rabi frequency equals the nuclear Zeeman frequency.

The activities of the joint NASA/DOE/DOD Nuclear Propulsion Program Technical Panels have been used as the basis for the current development of safety policies and requirements for the Space Exploration Initiatives (SEI) Nuclear Propulsion Technology development program. The Safety Division of the NASA Office of Safety and Mission Quality has initiated efforts to develop policies for the safe use of nuclear propulsion in space through involvement in the joint agency NuclearSafety Policy Working Group (NSPWG), encouraged expansion of the initial policy development into proposed programmatic requirements, and suggested further expansion into the overall risk assessment and risk management process for the NASA Exploration Program. Similar efforts are underway within the Department of Energy to ensure the safe development and testing of nuclear propulsion systems on Earth. This paper describes the NASA safety policy related to requirements for the design of systems that may operate where Earth re-entry is a possibility. The expected plan of action is to support and oversee activities related to the technology development of nuclear propulsion in space, and support the overall safety and risk management program being developed for the NASA Exploration Program.

We propose that in order to systematically improve healthcare quality, healthcare organizations (HCOs) need work environments that are person-centered: environments that support the careprovider as well as the patient. We further argue that HCOs have a moral imperative to provide a workplace where professional care standards can be achieved. We draw upon a large body of research from several disciplines to propose and articulate a theoretical framework that explains how the work environment should be related to the well-being of patients and careproviders, that is, the potential mediating mechanisms. Person-centered work environments include: 1. Climates for patient-centered care. 2. Climates for quality improvement. 3. Benevolent ethical climates. Such a work environment should support the provision of patient-centered care, and should lead to positive psychological states for careproviders, including psychological safety and positive affect. The model contributes to theory by specifying relationships between important organizational variables. The model can potentially contribute to practice by linking specific work environment attributes to outcomes for careproviders and patients. PMID:18839753

In an effort to mitigate the risks associated with satellite close approaches in the geostationary belt, satellite operators began to come together in early 2008 to establish a prototype GEO data center. That prototype provided a framework for operators to share orbital data for their fleets to be used to perform conjunction analysis and provide automated notifications of close approaches via the SOCRATES-GEO service. That service was extended to LEO operations in mid-2009 and, as of early 2010, the prototype was supporting 20 operators from over a dozen countries by automatically screening 300 satellites for close approaches twice each day. In April 2010, the prototype data center operated by the Center for Space Standards & Innovation (CSSI) was a key reason AGI was selected by the Space Data Association (SDA) to develop the SDA’s new Space Data Center (SDC). This paper will address how the SDC will use a service-oriented architecture (SOA) to support orbital operations by increasing the efficiency of analysis to mitigate the risk of conjunctions and radio frequency interference, thereby enhancing overall safety of flight.

Modern nuclear power technology still has significant sources of risk, and, weak links, such as, a threat of severe accidents with catastrophic unpredictable consequences and damage to the population, proliferation of nuclear weapon-usable materials, risks of long-term storage of toxic radioactive waste, risks of loss of major investments in nuclear facilities and their construction, lack of fuel resources for the ambitious role of nuclear power in the competitive balance of energy. Each of these risks is important and almost independent, though the elimination of some of them does not significantly alter the overall assessment of nuclear power.

The Defense NuclearSafeties Board (DNFSB) issued Recommendation 2000-2 on March 8, 2000, concerning the degrading conditions of vital safety systems, or systems important to nuclearsafety, at DOE sites across the nation. The Board recommended that the DOE take action to assess the condition of its nuclear systems to ensure continued operational readiness of vital safety systems that are important for safely accomplishing the DOE's mission. To verify the readiness of vital safety systems, a two-phased approach was established. Phase I consisted of a qualitative assessment to approved criteria of the defined vital safety systems by operating contractor personnel, overseen by Federal field office personnel. Based on Phase I Assessment results, vital safety systems with significant deficiencies would be further assessed in Phase II, a more extensive quantitative assessment, by a contractor and Federal team, using a second set of criteria. In addition, Defense Nuclear Facility Safety Board Recommendation 2000-2 concluded that the degradation of confinement ventilation systems was of major concern, and issued a separate set of criteria to perform a Phase II Assessment on confinement ventilation systems.

This Department of Energy (DOE) is approved for use by all components of DOE. It contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE Non-Reactor Nuclear Facilities. Adherence with these guidelines will provide consistency and uniformity in Criticality Safety Evaluations (CSEs) across the complex and will document compliance with DOE Order 5480.24 requirements as they pertain to CSEs.

The report strongly recommends that, with the end of the Cold War, safety and health at DOE facilities should be regulated by outside agencies rather than by any regulatory scheme, DOE must maintain a strong internal safety management system; essentially all aspects of safety at DOE`s nuclear facilities should be externally regulated; and existing agencies rather than a new one should be responsible for external regulation.

The report strongly recommends that, with the end of the Cold War, safety and health at DOE facilities should be regulated by outside agencies rather than by DOE itself. The three major recommendations are: under any regulatory scheme, DOE must maintain a strong internal safety management system; essentially all aspects of safety at DOE`s nuclear facilities should be externally regulated; and existing agencies rather than a new one should be responsible for external regulation.

Presents the summary section of the American Physical Society (APS) report on the safety features of the light-water reactor, reviews the design, construction, and operation of a reactor and outlines the primary engineered safety features. Summarizes the major recommendations of the study group. (GS)

Nuclear power plants are used as an option to meet the demands for electricity due to the low emission of CO2 and other contaminants. The accident at the Fukushima nuclear power plant in 2011 has forced the Chinese government to adjust its original plans for nuclear power. The construction of inland nuclear power plants was stopped, and construction is currently only permitted in coastal zones. However, one obstacle of those plants is that the elevation of those plants is notably low, ranging from 2 to 9 meters and a number of the nuclear power plants are located in or near geological fault zones. In addition, the population density is very high in the coastal zones of China. To reduce those risks of nuclear power plants, central government should close the nuclear power plants within the fault zones, evaluate the combined effects of storm surges, inland floods and tidal waves on nuclear power plants and build closed dams around nuclear power plants to prevent damage from storm surges and tidal waves. The areas without fault zones and with low elevation should be considered to be possible sites for future nuclear power plants if the elevation can be increased using soil or civil materials.

NASA formed the NASA Engineering and SafetyCenter in 2003 following the Space Shuttle Columbia accident. It is an Agency level, program-independent engineering resource supporting NASA's missions, programs, and projects. It functions to identify, resolve, and communicate engineering issues, risks, and, particularly, alternative technical opinions, to NASA senior management. The goal is to help ensure fully informed, risk-based programmatic and operational decision-making processes. To date, the NASA Engineering and SafetyCenter (NESC) has conducted or is actively working over 600 technical studies and projects, spread across all NASA Mission Directorates, and for various other U.S. Government and non-governmental agencies and organizations. Since inception, NESC human spaceflight related activities, in particular, have transitioned from Shuttle Return-to-Flight and completion of the International Space Station (ISS) to ISS operations and Orion Multi-purpose Crew Vehicle (MPCV), Space Launch System (SLS), and Commercial Crew Program (CCP) vehicle design, integration, test, and certification. This transition has changed the character of NESC studies. For these development programs, the NESC must operate in a broader, system-level design and certification context as compared to the reactive, time-critical, hardware specific nature of flight operations support.

Investigations performed in the last few years by the State Science Center of the Russian Federation - Academician A. N. Krylov Central Scientific-Research Institute, together with specialized enterprises of the Ministry of Atomic Energy of the Russian Federation, Sudprom, and other agencies of Russia, have shown the promise of marine nuclear power plants for producing underground nuclear power plants with a higher degree of protection from external and internal actions of different intensity. The concept was developed on the basis of an analysis of the energy supply in different regions of Russia and the near-abroad using fossil fuels (lignite, oil, natural gas). The change in the international environment, which makes it possible to convert the military technology, frees the industrial potential and skilled workers in Russia for development of products for the national economy. Stricter international standards and rules for increased safety and protection of nuclear power plants made it necessary to develop a new generation of reactors for ground-based power plants, which under the modern economic conditions cannot be implemented within the time periods acceptable for economics for most of the countries surrounding Russia. In the development of a new generation of ground-based nuclear power plants, the intense improvement of the aviation and space technology must be taken into account. This is connected with the increase in the catastrophes and the threat they present to the safety of unprotected power plants. This article is an abstract of the entire report.

...The U.S. Nuclear Regulatory Commission (NRC or the Commission) is issuing for public comment draft regulatory guide (DG), DG-1206, ``Configuration Management Plan for Digital Computer Software Used in Safety Systems of Nuclear Power Plants.'' The DG-1206 is proposed Revision 1 of RG 1.169, dated September 1997. This revision endorses, with clarifications, the enhanced consensus practices for......

... Inventory Analysis/FY 2012 Service Contract Inventory AGENCY: Defense Nuclear Facilities Safety Board (DNFSB... throughout the agency. The inventory has been developed in accordance with guidance issued on December 19... From the Federal Register Online via the Government Publishing Office DEFENSE NUCLEAR...

The paper presents a general review of public safety standards as adapted by the nuclear weapons testing program in the United States, and the impact of these changing standards on the nuclear testing program itself. The review notes the importance of improvements in diagnostic i...

The ASME NOG-1 standard, 'Rules for Construction of Overhead and Gantry Cranes', covers top running cranes for nuclear facilities; with the ASME NUM-1 standard, 'Rules for Construction of Cranes, Monorails, and Hoists', covering the single girder, underhung, wall and jib cranes, as well as the monorails and hoists. These two ASME nuclear crane standards provide criteria for designing, inspecting and testing overhead handling equipment with enhanced safety to meet the 'defense-in-depth' approach of the United States Nuclear Regulatory Commission (USNRC) documents NUREG 0554 and NUREG 0612. In addition to providing designs for enhanced safety, the ASME nuclear crane standards provide a basis for purchasing overhead handling equipment with standard safety features, based upon accepted engineering principles, and including performance and environmental parameters specific to nuclear facilities. The ASME NOG-1 and ASME NUM-1 standards not only provide enhanced safety for handling a critical load, but also increase profit by minimizing the possibility of load drops, by reducing cumbersome operating restrictions, and by providing the foundation for a sound licensing position. The ASME nuclear crane standards can also increase profit by providing the designs and information to help ensure that the right standard equipment is purchased. Additionally, the ASME nuclear crane standards can increase profit by providing designs and information to help address current issues, such as the qualification of nuclear plant cranes for making 'planned engineered lifts' for steam generator replacement and decommissioning.

A short History of the Joint Institute of the Nuclear Research in Dubna (Russian Federation) has been given. The Facilities : The Synchrophasotron, the Nuclotron-M, the IBR-2 Reactor,The Heavy Ions Isochronous cyclotrons U-400 and U-400M, Phasotron, The impuls resonance neutronic device IREN has been described. A short history of collaboration between JINR and scientists from Moldova has been given. The main new tendencies in Science management has been outlined.

The objective of the Safety Basis Implementation is to ensure that implementation of activities is accomplished in order to support readiness to move spent fuel from K West Basin. Activities may be performed directly by the Safety Basis Implementation Team or they may be performed by other organizations and tracked by the Team. This strategy will focus on five key elements, (1) Administration of Safety Basis Implementation (general items), (2) Implementing documents, (3) Implementing equipment (including verification of operability), (4) Training, (5) SNF Project Technical Requirements (STRS) database system.

Decisions regarding the verification of design plant lifetime and potential license renewal periods involve a determination of the component and circuit condition. In Service Inspection of key reactor components becomes a crucial consideration for continued safe plant operation. The determination of the equipment properties by Non Destructive Techniques during periodic intervals is an important aspect of the assessment of fitness-for-service and safe operation of nuclear power plants The Tacis and Phare were established since 1991 by the European Union as support mechanisms through which projects could be identified and addressed satisfactorily. In NuclearSafety, the countries mainly concerned are Russia, Ukraine, Armenia, and Kazakhstan for the Tacis programme, and Bulgaria, Czech Republic, Hungary, Slovak Republic, Lithuania, Romania and Slovenia for the Phare programme. The Tacis and Phare programs concerning the Nuclear Power Plants consist of: - On Site Assistance and Operational Safety, - Design Safety, - Regulatory Authorities, - Waste management, and are focused on reactor safety issues, contributing to the improvement in the safety of East European reactors and providing technology and safety culture transfer. The main parts of these programmes are related to the On-Site Assistance and to the Design Safety of VVER and RBMK Nuclear power plants where Non Destructive Techniques for In Service Inspection of the primary circuit components are addressed. (authors)

Organizational culture has a powerful influence on overall corporate performance. The ability to sustain superior results in ensuring the public`s health and safety is predicated on an organization`s deeply embedded values and behavioral norms and how these affect the ability to change and seek continuous improvement. The nuclear industry is developing increased recognition of the relationship of culture to nuclearsafety performance as a critical element of corporate strategy. This paper describes a self-assessment methodology designed to characterize and improve the nuclearsafety culture, including processes for addressing employee concerns. This methodology has been successfully applied on more than 30 occasions in the last several years, resulting in measurable improvements in safety performance and quality and employee motivation, productivity, and morale. Benefits and lessons learned are also presented.

The Engineering and Technology Office of Waste Processing identifies and reduces engineering and technical risks associated with key waste processing project decisions. The risks, and actions taken to mitigate those risks, are determined through technology readiness assessments, program reviews, technology information exchanges, external technical reviews, technical assistance, and targeted technology development and deployment (TDD). The Office of Waste Processing TDD program prioritizes and approves research and development scopes of work that address nuclearsafety related to processing of highly radioactive nuclear wastes. Thirteen of the thirty-five R&D approved work scopes in FY2009 relate directly to nuclearsafety, and are presented in this report.

In 1991, LLNL was asked by the NRC to provide technical assistance in various aspects of computer technology that apply to computer-based reactor protection systems. This has involved the review of safety aspects of new reactor designs and the provision of technical advice on the use of computer technology in systems important to reactor safety. The latter includes determining and documenting state-of-the-art subjects that require regulatory involvement by the NRC because of their importance in the development and implementation of digital computer safety systems. These subjects include data communications, formal methods, testing, software hazards analysis, verification and validation, computer security, performance, software complexity and others. One topic software reliability and safety is the subject of this paper.

As the current fleet of nuclear power plants in the US reaches 25+ years of operation, obsolescence is driving many utilities to implement upgrades to both their safety and non-safety-related Instrumentation and Control (I and C) Systems. Digital technology is the predominant replacement technology for these upgrades. Within the last 15 years, digital control systems have been deployed in non-safety- related control applications at many utilities. In addition, a few utilities have replaced small safety-related systems utilizing digital technology. These systems have shown digital technology to be robust, reliable and simpler to maintain. Based upon this success, acceptance of digital technology has gained momentum with both utilities and regulatory agencies. Today, in an effort to extend the operating lives of their nuclear stations and resolve obsolescence of critical components, utilities are now pursuing digital technology for replacement of their primary safety systems. AREVA is leading this effort in the United States with the first significant digital upgrade of a major safety system. AREVA has previously completed upgrades to safety-related control systems emergency diesel engine controls and governor control systems for a hydro station which serves as the emergency power source for a nuclear station. Currently, AREVA is implementing the replacement of both the Reactor Protection System (RPS) and the Engineered Safety Features Actuation System (ESFAS) on all three units at a US PWR site. (authors)

This paper describes the roles and functions of a special Nuclear Technology Center which is developed as an integral part of a permanent lunar base. Numerous contemporary studies clearly point out that nuclear energy technology will play a major role in any successful lunar/Mars initiative program and in the overall establishment of humanity's solar system civilization. The key role of nuclear energy in the providing power has been recognized. A Nuclear Technology Center developed as part of of a permanent lunar base can also help bring about many other nuclear technology applications, such as producing radioisotopes for self-illumination, food preservation, waste sterilization, and medical treatment; providing thermal energy for mining, materials processing and agricultural; and as a source of emergency habitat power. Designing such a center will involve the deployment, operation, servicing and waste product management and disposal of megawatt class reactor power plants. This challenge must be met with a minimum of direct human support at the facility. Furthermore, to support the timely, efficient integration of this Nuclear Technology Center in the evolving lunar base infrastructure, an analog of such a facility will be needed here on Earth. 12 refs., 4 figs., 1 tab.

This paper describes the roles and functions of a special Nuclear Technology Center which is developed as an integral part of a permanent lunar base. Numerous contemporary studies clearly point out that nuclear energy technology will play a major role in any successful lunar/Mars initiative program and in the overall establishment of humanity's solar system civilization. The key role of nuclear energy in the providing power has been recognized. A Nuclear Technology Center developed as part of a permanent lunar base can also help bring about many other nuclear technology applications, such as producing radioisotopes for self-illumination, food preservation, waste sterilization, and medical treatment; providing thermal energy for mining, materials processing and agricultural; and as a source of emergency habitat power. Designing such a center will involve the deployment, operation, servicing and waste product management and disposal of megawatt class reactor power plants. This challenge must be met with a minimum of direct human support at the facility. Furthermore, to support the timely, efficient integration of this Nuclear Technology Center in the evolving lunar base infrastructure, an analog of such a facility will be needed here on Earth.

Operations involving significant quantities of fissile material have been conducted at Los Alamos National Laboratory continuously since 1943. Until the advent of the Laboratory`s Nuclear Criticality Safety Committee (NCSC) in 1957, line management had sole responsibility for controlling criticality risks. From 1957 until 1961, the NCSC was the Laboratory body which promulgated policy guidance as well as some technical guidance for specific operations. In 1961 the Laboratory created the position of Nuclear Criticality Safety Office (in addition to the NCSC). In 1980, Laboratory management moved the Criticality Safety Officer (and one other LACEF staff member who, by that time, was also working nearly full-time on criticality safety issues) into the Health Division office. Later that same year the Criticality Safety Group, H-6 (at that time) was created within H-Division, and staffed by these two individuals. The training and education of these individuals in the art of criticality safety was almost entirely self-regulated, depending heavily on technical interactions between each other, as well as NCSC, LACEF, operations, other facility, and broader criticality safety community personnel. Although the Los Alamos criticality safety group has grown both in size and formality of operations since 1980, the basic philosophy that a criticality specialist must be developed through mentoring and self motivation remains the same. Formally, this philosophy has been captured in an internal policy, document ``Conduct of Business in the Nuclear Criticality Safety Group.`` There are no short cuts or substitutes in the development of a criticality safety specialist. A person must have a self-motivated personality, excellent communications skills, a thorough understanding of the principals of neutron physics, a safety-conscious and helpful attitude, a good perspective of real risk, as well as a detailed understanding of process operations and credible upsets.

Introduction: Professional radiation exposure cannot be avoided in nuclear medicine practices. It can only be minimized up to some extent by implementing good work practices. Aim and Objectives: The aim of our study was to audit the professional radiation exposure and exposure rate of radiation worker working in and around Department of nuclear medicine and molecular imaging, Tata Memorial Hospital. Materials and Methods: We calculated the total number of nuclear medicine and positron emission tomography/computed tomography (PET/CT) procedures performed in our department and the radiation exposure to the radiation professionals from year 2009 to 2012. Results: We performed an average of 6478 PET/CT scans and 3856 nuclear medicine scans/year from January 2009 to December 2012. The average annual whole body radiation exposure to nuclear medicine physician, technologist and nursing staff are 1.74 mSv, 2.93 mSv and 4.03 mSv respectively. Conclusion: Efficient management and deployment of personnel is of utmost importance to optimize radiation exposure in a high volume nuclear medicine setup in order to work without anxiety of high radiation exposure. PMID:25400361

The Secretary of Energy established independent Tiger Teams to conduct environment, safety, and health (ES H) compliance assessments at US Department of Energy (DOE) facilities. This report presents the assessment of the Feed Materials Production Center (FMPC) at Fernald, Ohio. The purpose of the assessment at FMPC is to provide the Secretary with information regarding current ES H compliance status, specific ES H noncompliance items, evaluation of the adequacy of the ES H organizations and resources (DOE and contractor), and root causes for noncompliance items. Areas reviewed included performance under Federal, state, and local agreements and permits; compliance with Federal, state and DOE orders and requirements; adequacy of operations and other site activities, such as training, procedures, document control, quality assurance, and emergency preparedness; and management and staff, including resources, planning, and interactions with outside agencies.

The radiation safety system (RSS) of the B-Factory accelerator facility at the Stanford Linear Accelerator Center (SLAC) is described. The RSS, which is designed to protect people from prompt radiation exposure due to beam operation, consists of the access control system (ACS) and the radiation containment system (RCS). The ACS prevents people from being exposed to the very high radiation levels inside a beamline shielding housing. The ACS consists of barriers, a standard entry module at every entrance, and beam stoppers. The RCS prevents people from being exposed to the radiation outside a shielding housing, due to either normal or abnormal operation. The RCS consists of power limiting devices, shielding, dump/collimator, and an active radiation monitor system. The inter-related system elements for the ACS and RCS, as well as the associated interlock network, are described. The policies and practices in setting up the RSS are also compared with the regulatory requirements.

The United States (US) Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of launching a Russian Topaz II space nuclear power system. A preliminary nuclearsafety assessment was conducted to determine whether or not a space mission could be conducted safely and within budget constraints. As part of this assessment, a safety policy and safety functional requirements were developed to guide both the safety assessment and future Topaz II activities. A review of the Russian flight safety program was conducted and documented. Our preliminary nuclearsafety assessment included a number of deterministic analyses, such as; neutronic analysis of normal and accident configurations, an evaluation of temperature coefficients of reactivity, a reentry and disposal analysis, an analysis of postulated launch abort impact accidents, and an analysis of postulated propellant fire and explosion accidents. Based on the assessment to date, it appears that it will be possible to safely launch the Topaz II system in the US with a modification to preclude water flooded criticality. A full scale safety program is now underway.

We have developed a new method to propagate the uncertainties of fundamental nuclear physics models and parameters used in nuclear data evaluation to the design and performance of future nuclear energy systems. Using Monte Carlo simulation, it is for the first time possible to couple these two fields at the extremes of nuclear science without any loss of information in between. With the help of a large database of nuclear reaction measurements, we have determined the uncertainties of theoretical nuclear reaction models such as the optical, compound nucleus, pre-equilibrium and fission models. A similar assessment is done for the parameters that describe the resolved resonance range. We are now able to quantify the required quality of theoretical nuclear reaction models and measurements directly from the reactor design requirements. Examples will be presented for actinides using criticality-safety benchmarks with feedback on experimental requirements.

The studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems, pertained to the General-Purpose Heat Source (compatibility and safety verification) and to the Light-Weight Radioisotope Heater units (overpressure and impact tests).

The nitrogen vacancy (NV) center in diamond is a paramagnetic defect with excellent spin properties that can reside within a few nanometers of the diamond surface, enabling atomic-scale magnetic resonance sensing of external nuclear spins. Here we use rotating frame longitudinal spin relaxation (T1ρ) based sensing schemes, known as Continuous Dynamical Decoupling (CDD), to detect external nuclear spins with shallow NV centers (<5 nm from the surface). Distinguishing neighboring nuclear spins from each other requires the NV center be near enough to create differences in the hyperfine shifts and coupling strengths of the nuclei. However, spin coherence time and consequently the sensitivity of dynamical decoupling techniques degrade sharply as NVs become shallower. We use strong continuous driving to overcome this fast decoherence and detect an ensemble of external nuclear spins using a single shallow NV center with a short T2 (<2 μs) at magnetic fields as high as 0.5 Tesla. The increased sensitivity of this method relative to pulsed dynamical decoupling techniques demonstrates the benefits of CDD for sensing with very shallow NV centers. This work was supported by DARPA, AFOSR, and the DIAMANT program.

The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel

Background Novel research has revealed that the relative risk of death increased by 10% and 15% for admissions on a Saturday and Sunday, respectively. With an imminent threat of 7-day services in the National Health Service, including weekend operating lists, handover plays a pivotal role in ensuring patient safety is paramount. This audit evaluated the quality, efficiency, and safety of surgical handover of pre- and postoperative cardiothoracic patients in a tertiary center against guidance on Safe Handover published by the Royal College of Surgeons of England and the British Medical Association. Methods A 16-item questionnaire prospectively audited the nature, time and duration of handover, patient details, operative history and current clinical status, interruptions during handover, and difficulties cross-covering specialties over a month. Results Just over half (52%) of the time, no handover took place. The majority of handovers (64%) occurred over the phone; two-thirds of these were uninterrupted. All handovers were less than 10 minutes in duration. About half of the time, the senior house officer had previously met the registrar involved in the handover, but the overwhelming majority felt it would facilitate the handover process if they had prior contact. Patient details handed over 100% of the time included name, ward, and current clinical diagnosis. A third of the time, the patient’s age, responsible consultant, and recent operations or procedures were not handed over, potentially compromising future management due to delays and lack of relevant information. Perhaps the most revealing result was that the overall safety of handover was perceived to be five out of ten, with ten being very safe with no aspects felt to impact negatively on optimal patient care. Conclusion These findings were presented to the department, and a handover proforma was implemented. Recommendations included the need for a new face-to-face handover. A reaudit will evaluate the

Available in abstract form only. Full text of publication follows: The French Atomic Energy Commission (CEA) center at Fontenay aux Roses (CEN-FAR) is the Commission's oldest center is located in the southern suburbs of Paris. It was opened on 26 March 1946 to host the first French nuclear reactor ZOE that went critical on 12 December 1946. The first laboratories were installed in existing buildings on the site. (authors)

Safety practices in health care have not kept pace with the increasing complexity of medical technology. Although anaesthesia is generally considered to be a leader in the improvement of patient safety, more powerful safety strategies must be found and employed. From an analysis of system characteristics, the nuclear power industry is proposed as an alternative analogy for safety in anaesthesia, and a novel diagrammatic approach is developed for the conceptualisation of safety goals. The nuclear power industry has spent vastly more time and money than has health care on the development of safety, and has progressed through significant safety milestones approximately three times more quickly than has anaesthesia. The greatest scope for the improvement of safety in anaesthesia lies in the appropriate re-design of medical systems and the lowering of the threshold for the reporting of incidents to include accident precursors, thus allowing the identification of dangerous systems before accidents occur. PMID:16229697

In June of this year, the Department of Energy (DOE) issued directives DOE O 414.1C and DOE G 414.1-4 to improve quality assurance programs, processes, and procedures among its safety contractors. Specifically, guidance entitled, ''Safety Software Guide for use with 10 CFR 830 Subpart A, Quality Assurance Requirements, and DOE O 414.1C, Quality Assurance, DOE G 414.1-4'', provides information and acceptable methods to comply with safety software quality assurance (SQA) requirements. The guidance provides a roadmap for meeting DOE O 414.1C, ''Quality Assurance'', and the quality assurance program (QAP) requirements of Title 10 Code of Federal Regulations (CFR) 830, Subpart A, Quality Assurance, for DOE nuclear facilities and software application activities. [1, 2] The order and guide are part of a comprehensive implementation plan that addresses issues and concerns documented in Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2002-1. [3] Safety SQA requirements for DOE as well as National Nuclear Security Administration contractors are necessary to implement effective quality assurance (QA) processes and achieve safe nuclear facility operations. DOE G 414.1-4 was developed to provide guidance on establishing and implementing effective QA processes tied specifically to nuclear facility safety software applications. The Guide includes software application practices covered by appropriate national and international consensus standards and various processes currently in use at DOE facilities. While the safety software guidance is considered to be of sufficient rigor and depth to ensure acceptable reliability of safety software at all DOE nuclear facilities, new nuclear facilities are well suited to take advantage of the guide to ensure compliant programs and processes are implemented. Attributes such as the facility life-cycle stage and the hazardous nature of each facility operations are considered, along with the category and level of importance of the

Purpose: Over the last decade, the use of oral chemotherapy (OC) for the treatment of cancer has dramatically increased. Despite their route of administration, OCs pose many of the same risks as intravenous agents. In this quality improvement project, we sought to examine our current process for the prescription of OC at the Abramson Cancer Center of the University of Pennsylvania and to improve on its safety. Methods: A multidisciplinary team that included oncologists, advanced-practice providers, and pharmacists was formed to analyze the current state of our OC practice. Using Lean Six Sigma quality improvement tools, we identified a lack of pharmacist review of the OC prescription as an area for improvement. To address these deficiencies, we used our electronic medical system to route OC orders placed by treating providers to an oncology-specific outpatient pharmacist at the Abramson Cancer Center for review. Results: Over 7 months, 63 orders for OC were placed for 45 individual patients. Of the 63 orders, all were reviewed by pharmacists, and, as a result, 22 interventions were made (35%). Types of interventions included dosage adjustment (one of 22), identification of an interacting drug (nine of 22), and recommendations for additional drug monitoring (12 of 22). Conclusion: OC poses many of the same risks as intravenous chemotherapy and should be prescribed and reviewed with the same oversight. At our institution, involvement of an oncology-trained pharmacist in the review of OC led to meaningful interventions in one third of the orders. PMID:26733627

Four slides present a historical perspective on the evolution of nuclearsafety, a description of systemic misalignment (available resources do not match expectations, demographic cliff developing, promulgation of increased expectations and new requirements proceeds unabated), and needs facing nuclearsafety (financial stability, operational stability, and succession planning). The following conclusions are stated under the heading "NuclearSafety - 'The System'": the current universe of requirements is too large for the resource pool available; the current universe of requirements has too many different sources of interpretation; there are so many indicators that it’s hard to know what is leading (or important); and the net result can come to defy integrated comprehension at the worker level.

Safety compliance is of paramount importance in guaranteeing the safe running of nuclear power plants. However, it depends mostly on procedures that do not always involve the safest outcomes. This article introduces an empirical model based on the organizational role theory to analyze the influence of legitimate sources of expectations (procedures formalization and leadership) on workers' compliance behaviors. The sample was composed of 495 employees from two Spanish nuclear power plants. Structural equation analysis showed that, in spite of some problematic effects of proceduralization (such as role conflict and role ambiguity), procedure formalization along with an empowering leadership style lead to safety compliance by clarifying a worker's role in safety. Implications of these findings for safety research are outlined, as well as their practical implications. PMID:24495145

In the framework of the preparation for the decommissioning of the Ignalina Nuclear Power Plant (INPP) a new Interim Spent Nuclear Fuel Storage Facility (ISFSF) will be built in the existing sanitary protection zone (SPZ) of INPP. In addition to the ISFSF, the new spent nuclear fuel management activity will include all necessary spent nuclear fuel retrieval and packaging operations at the Reactor Units, transfer of storage casks to the ISFSF, and other activities appropriate to the chosen design solution and required for the safe removal of the existing spent nuclear fuel from storage pools and insertion into the new ISFSF. The Republic of Lithuania regulations require that the average annual dose to the critical group members of population due to operation of nuclear facility shall not exceed dose constraint. If several nuclear facilities are located in the same SPZ, the same dose constraint shall envelope radiological impacts from all operating and planned nuclear facilities. The paper discusses radiological safety assessment aspects as relevant for the new nuclear activity to be implemented in the SPZ of INPP considering specificity of Lithuanian regulatory requirements. The safety assessment methodology aspects, results and conclusions as concern public exposure are outlined and discussed. (authors)

Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based storage and dry transport containers under various loading patterns and moderating conditions. To comply with standard safety requirements, fresh 4.25% enriched nuclear fuel was assumed. The impact of potential optimum moderation due to water steam or foam formation as well as of different interpretations, of neutron multiplication through varying the system boundary conditions was elaborated. The simulations indicate that in the case of compact (all rack locations filled with fresh fuel) single or 'double tiering' loading, the supercriticality can occur under the conditions of enhanced neutron moderation, due to accidentally reduced density of cooling water. Under standard operational conditions the effective multiplication factor (k{sub eff}) of pool-based storage facility remains below the specified safety limit of 0.95. The nuclearsafety requirements are fulfilled even when the fuel elements are arranged at a minimal distance, which can be initiated, for example, by an earthquake. The dry container in its recommended loading scheme with 26 fuel elements represents a safe alternative for the repository of fresh fuel. Even in the case of complete water flooding, the k{sub eff} remains below the specified safety level of 0.98. The criticality safety limit may however be exceeded with larger amounts of loaded fuel assemblies (i.e., 32). Additional Monte Carlo criticality safety analyses are scheduled to consider the 'burnup credit' of PWR spent nuclear fuel, based on the ongoing calculation of typical burnup activities.

All nuclear medicine nurses need to have some knowledge and awareness on radiation safety. At present, there is no study to address this issue in Malaysia. The aims of this study were (1) to determine the level of knowledge and awareness on radiation safety among nuclear medicine nurses at Putrajaya Hospital in Malaysia and (2) to assess the effectiveness of a training program provided by the hospital to increase the knowledge and awareness of the nuclear medicine nurses. A total of 27 respondents attending a training program on radiation safety were asked to complete a questionnaire. The questionnaire consists 16 items and were categorized into two main areas, namely general radiation knowledge and radiation safety. Survey data were collected before and after the training and were analyzed using descriptive statistics and paired sample t-test. Respondents were scored out of a total of 16 marks with 8 marks for each area. The findings showed that the range of total scores obtained by the nuclear medicine nurses before and after the training were 6-14 (with a mean score of 11.19) and 13-16 marks (with a mean score of 14.85), respectively. Findings also revealed that the mean score for the area of general radiation knowledge (7.59) was higher than that of the radiation safety (7.26). Currently, the knowledge and awareness on radiation safety among the nuclear medicine nurses are at the moderate level. It is recommended that a national study be conducted to assess and increase the level of knowledge and awareness among all nuclear medicine nurses in Malaysia.

To examine the feasibility of establishing Highway Safety Manpower Development and Research (HSMDR) Centers at university-level institutions which would produce three types of manpower--safety specialists, safety professionals, and research manpower, previous National Highway Safety Bureau research studies and approximately 50 federally funded…

Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

Probabilistic risk assessment (PRA) has proven to be an important tool in the safety assessment of nuclear reactors throughout the world. Decision making with regard to many safety issues has been facilitated by both general insights from and direct application of this technology. Key uses of PRA are discussed and some examples of successful applications are cited. The benefits and limitations of PRA are also discussed as well as the broader outlook for applications of PRA. 9 refs.

This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed are ongoing; the results and conclusions described may change as the work continues. 36 figs.

This quarterly report describes studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems, carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed are ongoing; the results and conclusions described may change as the work progresses.

This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed here are ongoing. Results and conclusions described may change as the work continues.

This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory.

This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed here are ongoing. Results and conclusions described may change as the work continues.

This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed here are ongoing. Results and conclusions described may change as the work continues.

This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed here are ongoing. Results and conclusions may change as the work continues.

This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed here are ongoing. Results and conclusions described may change as the work continues.

This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems conducted for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed are ongoing; the results and conclusions described may change as the work progresses. 15 figs.

This quarterly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed are ongoing; the results and conclusions described may change as the work progresses.

This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed here are ongoing. Results and conclusions described may change as the work continues.

This quarterly report describes studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems, which were carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed are ongoing; the results and conclusions described may change as the work progresses.

This quarterly report describes studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems, carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. The studies discussed are ongoing; the results and conclusions described may change as the work progresses. 20 figs., 4 tabs.

Ukraine operates thirteen (13) Soviet-designed pressurized water reactors, VVERS. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs), in accordance with new SAR content requirements issued in September 1995, by the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine. The requirements are in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. The last two requirements, on PRA and BDBA, are new, and the DBA requirements are an expanded version of the older SAR requirements. The US Department of Energy (USDOE), as part of its Soviet-Designed Reactor Safety activities, is providing assistance and technology transfer to Ukraine to support their nuclear power plants (NPPs) in developing a Western-type technical basis for the new SARs. USDOE sponsored In-Depth Safety Assessments (ISAs) are in progress at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1, and a follow-on study has been initiated at Khmenytskyy Unit 1. The ISA projects encompass most areas of plant safety evaluation, but the initial emphasis is on performing a detailed, plant-specific Level 1 Internal Events PRA. This allows the early definition of the plant risk profile, the identification of risk significant accident sequences and plant vulnerabilities and provides guidance for the remainder of the safety assessments.

This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains - in chronological order - the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41.

In August 1986, after 6 years of effort, the U.S. Nuclear Regulatory Commission adopted a Policy Statement on safety goals for nuclear power reactors. The commission's qualitative goals state that individual members of the public should be provided a level of protection from the consequences of nuclear power plant operation such that they bear no significant additional risk to life and health, and societal risks to life and health from nuclear power should be comparable to or less than the risks of generating electricity by viable competing technologies and should not be a significant addition to other societal risks. The commission's safety goal Policy Statement also includes quantitative design objectives. PMID:3563510

This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.

Computerized fault tree analyses are used to obtain both qualitative and quantitative information about the safety and reliability of an electrical control system that shuts the reactor down when certain safety criteria are exceeded, in the design of a nuclear plant protection system, and in an investigation of a backup emergency system for reactor shutdown. The fault tree yields the modes by which the system failure or accident will occur, the most critical failure or accident causing areas, detailed failure probabilities, and the response of safety or reliability to design modifications and maintenance schemes.

The occurrence of certain potential events in nuclear fuel reprocessing plants could lead to significant consequences involving risk to operating personnel or to the general public. This document is a compilation of such potential initiating events in nuclear fuel reprocessing plants. Possible general incidents and incidents specific to key operations in fuel reprocessing are considered, including possible causes, consequences, and safety features designed to prevent, detect, or mitigate such incidents.

This paper summarizes the presentations and the insights offered by panelists John P. Winston, Robert Bernero, and Stephen LaMontagne during the Panel on Safety, Health and the Environment: Implications of Nuclear Power Growth that took place during the NCRP 2009 Annual Meeting. The paper describes the opportunities and the challenges faced in the areas of infrastructure development, radiation control, licensing and regulatory issues, and non-proliferation as a consequence of the forecasted growth in nuclear power capacity worldwide. PMID:21399405

A building project of Radiotherapy & Nuclear Medicine services (diagnostic and therapy), within an Integral Oncology Center (IOC), requires interdisciplinary participation of architects, biomedical engineers, radiation oncologists and medical physicists. This report focus on the medical physicist role in designing, building and commissioning stages, for the final clinical use of an IOC at the Oaxaca High Specialization Regional Hospital (HRAEO). As a first step, during design stage, the medical physicist participates in discussions about radiation safety and regulatory requirements for the National Regulatory Agency (called CNSNS in Mexico). Medical physicists propose solutions to clinical needs and take decisions about installing medical equipment, in order to fulfill technical and medical requirements. As a second step, during the construction stage, medical physicists keep an eye on building materials and structural specifications. Meanwhile, regulatory documentation must be sent to CNSNS. This documentation compiles information about medical equipment, radioactivity facility, radiation workers and nuclear material data, in order to obtain the license for the linear accelerator, brachytherapy and nuclear medicine facilities. As a final step, after equipment installation, the commissioning stage takes place. As the conclusion, we show that medical physicists are essentials in order to fulfill with Mexican regulatory requirements in medical facilities.

To develop safe nuclear energy production acceptable to the society it is proposed to use in the future strongly subcritical reactors (k=0.96-0.97) driven by proton or deuteron accelerators. The accelerator with the current of 40mA and particle energy {approximately}0.8 GeV/nucleon will provide 2 GW (th.) reactor power in fast reactor with metallic U-Pu fuel. The design, control and parameters of such a system are discussed.

This publication brings together, in concise form, information regarding the many recommendations and requirements for safe operation of a nuclear medicine laboratory. The need for such a compendium was perceived by the staff of the Nuclear Medicine Laboratory. This need arises from several sources. Many individuals enter the field with little training in the handling of radioactive materials; for example, a physician trained in cardiology, oncology, or neurology. The increasing development of portable instrumentation has allowed movement of radiopharmaceuticals from the confines of the nuclear medicine lab to coronary and intensive care facilities where personnel may lack adequate knowledge of safe handling procedures. A health physicist, trained to account for all radioactive material placed under his control, may have difficulty adapting to the accepted practice of releasing a patient who has been administered millicurie quantities of radioactivity, with little or no control over subsequent disposal of excreta. Further differences exist between handling practices for radioactive materials in the scientific laboratory and in the medical facility. This guide tries where possible to clarify some of these issues.

Dimethyl fumarate (DMF) was recently approved for treating patients with relapsing-remitting multiple sclerosis (RRMS) based on two phase III clinical trials demonstrating its efficacy. This prompts the need for demonstrating the clinical efficacy and safety of DMF in the real world. By retrospective analysis of medical records at two German MS centers, 644 MS patients treated with DMF were identified. All were included in a safety analysis, and a subgroup of patients with available efficacy data during previous MS therapies (n = 352) was further analyzed for annualized relapse rate and disability progression assessed by the EDSS. In the overall DMF population studied, the annualized relapse rate decreased from 0.52 at baseline to 0.35, and the annualized disability progression from 0.15 to 0.10. Patients who were switched from interferons or glatiramer acetate to DMF revealed a greater benefit, whereas patients pretreated with more potent immunotherapies did not respond that well. Interestingly, patients with a lymphocyte count ≥2000/µl after 0.52 years (mean, SD 0.2) of DMF treatment did not benefit compared to those with lower lymphocyte counts. In total, 22.2 % of the patients withdrew from DMF due to side effects, with gastrointestinal discomfort (12.7 %) and lymphopenia (5.3 %) as most frequently reported reasons. Our study corroborates that DMF is an overall safe and effective drug that reduces relapse rate as well as disability progression in MS patients. Further prospective studies are warranted to establish the additional parameters predicting DMF response, especially in patients switching from other first-line immunotherapies. PMID:27260297

The Democratic People's Republic of Korea (DPRK) announced the conduct of a third underground nuclear test on 12 February 2013 in the northeastern part of the country as the previous tests that were conducted in 2009 and 2006. The latest nuclear test is the best detected nuclear event by the global seismic networks. The magnitude estimates show that each new test increased in size when compared with the previous one. As Turkish NDC (National Data Center), we have analyzed the 2013 and 2009 nuclear tests using seismic data from International Monitoring System (IMS) stations through the International Data Center (IDC) located in Vienna. Discrimination analysis was performed based on mb:Ms magnitude ratio and spectral analysis. We have also applied array based waveform cross-correlation to show the similarity of the nuclear tests and precise arrival time measurements for relative location estimates and basic infrasound analysis using two IMS infrasound stations for the 2013 event. Seismic analysis were performed using softwares such as Geotool, EP (Event processor from Norsar) and Seismic Analysis Code (SAC) and the infrasound data were analyzed by using PMCC from CEA-France. The IMS network is operating under the Comprehensive Nuclear Test Ban Treaty Organization (CTBTO). The CTBTO verification system is under continuous development, also making use of the state of the art technologies and methodologies.

Priorities and achievements of the Kennedy Space Center (KSF) Instrumentation Laboratories in improving operational safety and decreasing processing costs associated with the Shuttle vehicle are addressed. Technologies that have been or are in the process of technology transfer are reviewed, and routes by which commercial concerns can obtain licenses to other KSF Instrumentation Laboratory technologies are discussed.

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Scientific Successes • The first phonon density of states (PDOS) measurements for UO2 to include anharmonicity were obtained using time-of-flight inelastic neutron scattering at the Spallation Neutron Source (SNS), and an innovative, experimental-based anharmonic smoothing technique has enabled quantitative benchmarking of ab initio PDOS simulations. • Direct comparison between anharmonicity-smoothed ab initio PDOS simulations for UO2 and experimental measurements has demonstrated the need for improved understanding of UO2 at the level of phonon dispersion, and, further, that advanced lattice dynamics simulations including finite temperatures approaches will be required for handling this strongly correlated nuclear fuel. • PDOS measurements performed on polycrystalline samples have identified the phonon branches and energy ranges most highly impacted by fission-product and hyper-stoichiometry lattice defects in UO2. These measurements have revealed the broad-spectrum impact of oxygen hyper-stoichiometry on thermal transport. The reduction in thermal conductivity caused by hyper-stoichiometry is many times stronger than that caused by substitutional fission-product impurities. • Laser-based thermo-reflectance measurements on UO2 samples irradiated with light (i.e. He) ions to introduce point defects have been coupled with MD simulations and lattice parameter measurements to determine the role of uranium and oxygen point defects in reducing thermal conductivity. • A rigorous perturbation theory treatment of phonon lifetimes in UO2 based on a 3D discretization of the Brillouin zone coupled with experimentally measured phonon dispersion has been implemented that produces improved predictions of the temperature dependent thermal conductivity. • Atom probe investigations of the influence of grain boundary structure on the segregation behavior of Kr in UO2 have shown that smaller amounts of Kr are present at low angle grain boundaries than at large angle grain

Review of NMP-NCS-94-0087, ``Nuclear Criticality Safety Evaluation 94-02: Uranium Solidification Facility Pencil Tank Module Spacing (U), April 18, 1994,`` was requested of the SRTC Applied Physics Group. The NCSE is a criticality assessment to show that the USF process module spacing, as given in Non-Conformance Report SHM-0045, remains safe for operation. The NCSE under review concludes that the module spacing as given in Non-Conformance Report SHM-0045 remains in a critically safe configuration for all normal and single credible abnormal conditions. After a thorough review of the NCSE, this reviewer agrees with that conclusion.

The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.

The purpose of this article is to address common occupational hazards and safety concerns of nuclear medicine technologists. There are many possible occupational hazards, but this review is intended to concentrate on common hazards and safety concerns. These include radiation safety issues and concerns about the possibility of developing latent diseases, such as eye cataracts or cancer; pregnant workers and radiation safety issues; biohazard concerns associated with patient body fluids; possible low-back pain from moving heavy equipment and performing patient transfers; and possible repetitive trauma disorders, such as carpal tunnel syndrome, from computer work. Suggestions are made regarding how to identify potential hazards and avoid them. After reading this article, nuclear medicine technologists should be able to explain the importance of the as-low-as-reasonably-achievable concept, discuss the possible effects of ionizing radiation on the adult and the developing fetus, list several basic principles to avoid injury to the back, list and describe the more common repetitive trauma disorders or injuries and how to avoid them, and list and describe the biohazard safety issues that nuclear medicine technologists face and how to develop policy to minimize exposure risk. PMID:18287195

This white paper outlines the relevant regulatory policy and guidance for a risk-informed approach for establishing the safety classification of Structures, Systems, and Components (SSCs) for the Next Generation Nuclear Plant and sets forth certain facts for review and discussion in order facilitate an effective submittal leading to an NGNP Combined Operating License application under 10 CFR 52.

This report of the Biomedical and Environmental Effects Subpanel (BEES) of the Interagency NuclearSafety Review Panel (INSRP), for the Galileo space mission addresses the possible radiological consequences of postulated accidents that release radioactivity into the environment. This report presents estimates of the consequences and uncertainties given that the source term is released into the environment. 10 refs., 6 tabs.

... Opportunity for a Hearing,'' see 78 FR 16,876, 16,883 (Mar. 19, 2013), a hearing request was filed on May 20... in August 2007. See 72 FR 49,139. Issued at Rockville, Maryland this 28th day of May 2013. E. Roy... COMMISSION FirstEnergy Nuclear Operating Company; Establishment of Atomic Safety and Licensing Board...

... COMMISSION Southern Nuclear Operating Company Establishment of Atomic Safety And Licensing Board Pursuant to delegation by the Commission dated December 29, 1972, published in the Federal Register, 37 FR 28,710 (1972... accordance with the NRC E-Filing rule, which the NRC promulgated in August 2007 (72 FR 49,139). Issued...

The proposed paper reports the results of a preliminary investigation on safety impact deriving from the coupling of a desalination plant with a 600 MWe Passive Design PWR like the AP600 Nuclear Power Plant. This evaluation was performed in the frame of the EURODESAL Project of the 5. EURATOM Framework Programme. (authors)

Present nuclear rocket reactors under test in Nevada are operated at nominal power levels of 1000 Mw. It does not seem unreasonable in the future to anticipate reactors with power levels in the range up to 5,000 Mw for space applications. It has been shown that the normal testing of large nuclear rocket engines at NRDS could impose some restrictions on the fuel performance which would not otherwise be required by space flight operation. The only apparent alternative would require a capability for decontaminating effluent gases prior to release to the atmosphere. In addition to the source restrictions, tests will almost certainly be controlled by wind and atmospheric stability conditions, and the requirements for monitoring and control of off-site exposures will be much more stringent than those presently in force. An analysis of maximum accidents indicates that projections of present credible occurrences cannot be tolerated in larger engine tests. The apparent alternatives to a significant (order of magnitude or better) reduction in credible accident consequences, are the establishment of an underground test facility, a facility in an area equivalent to the Pacific weapons proving ground, or in space.

... identification as Draft Regulatory Guide, DG-1208 on August 22, 2012 (77 FR 50722) for a 60-day public comment... COMMISSION Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear Power Plants..., ``Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear Power Plants.''...

Objectives. We assessed the prevalence of regulatory noncompliance of licensed child care centers and identified factors associated with improved compliance. Methods. We analyzed 676 routine, unannounced reports of child care centers collected by the Connecticut Department of Public Health licensing specialists over a 2-year time period, included characteristics of centers, and created categories of regulations. Results. The sample included 41% of licensed child care centers. Of the 13 categories of regulations in the analyses, 7 categories (outdoor safety, indoor safety, indoor health, child and staff documentation, emergency preparedness, infant-toddler indoor health, and infant-toddler indoor safety) had regulations with center noncompliance greater than 10%. Playground hazard-free was the regulation with the highest frequency (48.4%) of noncompliance. Compliance with the regulation for 20 hours of continuing education per year for child care providers was the characteristic most frequently associated with regulations compliance. Conclusions. Efforts to support continuing education of child care providers are essential to improve and sustain healthy and safe early-care and education programs. Analyses of state child care licensing inspection reports provide valuable data and findings for strategic planning efforts. PMID:23948016

A comprehensive review of injuries sustained by personnel working on naval flight decks between January 1977 and December 1991 was conducted using database records maintained at the U.S. Naval SafetyCenter, Norfolk, VA. Data included all fatalities, permanent total disabilities, permanent partial disabilities, and major injuries resulting in 5 or more lost work days. Injuries were coded using ICD-9-CM codes for analysis. A total of 918 flight deck personnel were reported injured during this 15-yr period, including 43 fatalities, 5 permanent total disabilities, 42 permanent partial disabilities, and 828 major injuries. Of the non-fatalities, a plethora of fractures, traumatic amputations, major lacerations, dislocations, contusions, concussions, burns, crushing injuries, sprains, and strains were reported. Nearly all naval platforms with a flight deck reported an injury. While an average of 51 injuries per 100,000 aircraft recoveries were reported annually on aircraft carriers from 1977-86, a marked reduction to a rate of roughly 30 injuries was observed annually from 1987-90. What makes injuries sustained on the flight deck particularly disconcerting is that over 90% can be attributed to human causal factors. PMID:7646412

As part of the Interagency NuclearSafety Review Panel's assessment of the nuclearsafety of NASA's Ulysses Mission to investigate properties of the sun, the Power System Subpanel has reviewed the safety analyses and risk evaluations done for the General Purpose Heat Source-Radioisotope Thermoelectric Generator which provides on-board electrical power for the spacecraft. This paper summarizes the activities and results of that review. In general, the approach taken in the primary analysis, executed by the General Electric Company under contract to the Department of Energy, and the resulting conclusions were confirmed by the review. However, the Subpanel took some exceptions and modified the calculations accordingly, producing an independent evaluation of potential releases of radioactive fuel in launch and reentry accidents. Some of the more important of these exceptions are described briefly.

This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. A model is introduced for the purposes of organizing the literature review and showing key relationships among identified organizational factors and nuclear power plant safety. Volume I of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety.

Chemical and metallurgical operations involving plutonium and other nuclear materials account for most activities performed at the LANL's Plutonium Facility (PF-4). The presence of large quantities of fissile materials in numerous forms at PF-4 makes it necessary to maintain an active criticality safety program. The LANL Nuclear Criticality Safety (NCS) Program provides guidance to enable efficient operations while ensuring prevention of criticality accidents in the handling, storing, processing and transportation of fissionable material at PF-4. In order to achieve and sustain lower criticality safety control infraction (CSCI) rates, PF-4 operations are continuously improved, through the use of Lean Manufacturing and Six Sigma (LSS) business practices. Employing LSS, statistically significant variations (trends) can be identified in PF-4 CSCI reports. In this study, trends have been identified in the NCS Program using the NCS Database. An output metric has been developed that measures ADPSM Management progress toward meeting its NCS objectives and goals. Using a Pareto Chart, the primary CSCI attributes have been determined in order of those requiring the most management support. Data generated from analysis of CSCI data help identify and reduce number of corresponding attributes. In-field monitoring of CSCI's contribute to an organization's scientific and technological excellence by providing information that can be used to improve criticality safety operation safety. This increases technical knowledge and augments operational safety.

Chemical and metallurgical operations involving plutonium and other nuclear materials account for most activities performed at the LANL's Plutonium Facility (PF-4). The presence of large quantities of fissile materials in numerous forms at PF-4 makes it necessary to maintain an active criticality safety program. The LANL Nuclear Criticality Safety (NCS) Program provides guidance to enable efficient operations while ensuring prevention of criticality accidents in the handling, storing, processing and transportation of fissionable material at PF-4. In order to achieve and sustain lower criticality safety control infraction (CSCI) rates, PF-4 operations are continuously improved, through the use of Lean Manufacturing andmore » Six Sigma (LSS) business practices. Employing LSS, statistically significant variations (trends) can be identified in PF-4 CSCI reports. In this study, trends have been identified in the NCS Program using the NCS Database. An output metric has been developed that measures ADPSM Management progress toward meeting its NCS objectives and goals. Using a Pareto Chart, the primary CSCI attributes have been determined in order of those requiring the most management support. Data generated from analysis of CSCI data help identify and reduce number of corresponding attributes. In-field monitoring of CSCI's contribute to an organization's scientific and technological excellence by providing information that can be used to improve criticality safety operation safety. This increases technical knowledge and augments operational safety.« less

In recent years, the management of patient doses in medical imaging has evolved as concern about radiation exposure has increased. Efforts and techniques to reduce radiation doses are focussed not only on the basis of patient safety, but also on the fundamentals of justification and optimisation in cooperation with international organisations such as the International Commission on Radiological Protection, the International Atomic Energy Agency, and the World Health Organization. The Image Gently campaign in children and Image Wisely campaign in adults to lower radiation doses have been initiated in the USA. The European Association of Nuclear Medicine paediatric dosage card, North American consensus guidelines, and Nuclear Medicine Global Initiative have recommended the activities of radiopharmaceuticals that should be administered in children. Diagnostic reference levels (DRLs), developed predominantly in Europe, may be an important tool to manage patient doses. In Korea, overexposure to radiation, even from the use of medical imaging, has become a public issue, particularly since the accident at the Fukushima nuclear power plant. As a result, the Korean NuclearSafety and Security Commission revised the technical standards for radiation safety management in medical fields. In parallel, DRLs for nuclear medicine procedures have been collected on a nationwide scale. Notice of total effective dose from positron emission tomography-computed tomography for cancer screening has been mandatory since mid-November 2014. PMID:26960820

The Nuclear Science References (NSR) database is an indexed bibiliography of primary and secondary references in nuclear physics research. About 80 journals are regularly scanned for articles. Recent references are added on a weekly basis. Approximately 4300 entries are added to the database annually. In general, articles are included in NSR if they include measured, calculated, or deduced quantitative nuclear structure or reaction data. Papers that apply previously known data are generally not included. Examples of this include neutron activation analysis using known cross sections or radiological dating using known half-lives. The database can be searched like a normal bibliographic database but can also be searched by the data that distinguishes it, data such as the nuclide, target/parent/daughter, reaction, incident particles, and outgoing particles. (Specialized Interface) [Taken from the NSR Help pages at http://www.nndc.bnl.gov/nsr/nsr_help.jsp

... COMMISSION Atomic Safety and Licensing Board; In the Matter of Nuclear Innovation North America LLC (South... Statements) On October 31, 2011, the Atomic Safety and Licensing Board will convene an evidentiary hearing to... Daylight Time (EDT) on Monday, October 31, 2011, at the Atomic Safety and Licensing Board Panel...

Slovenia has plans to build a repository for low- and medium-radioactive waste by 2013, the location in the very neighborhood of nuclear power plant is almost chosen, but the final approval hasn't been granted yet. The main obstacle is public opinion. Public information activities are therefore vitally important. One of the most important players in this area in Slovenia is Nuclear Training Center in Ljubljana. Though its main task is training of nuclear professionals, it has a significant role in dissemination of knowledge about radioactivity and nuclear technology also among general public. Public information is focused on youngsters. Almost one half of every generation of schoolchildren in Slovenia visits the Information center yearly and in May 2007, we have celebrated the 100,000. visitor since its opening. Live lectures, exhibition, publications and laboratory demonstrations are offered. To measure the opinion of youngsters about nuclear power and get a feed-back for our activities about 1000 youngsters are polled every year since 1993 using the same basic set of questions. The paper describes the information strategy, types of lectures and information materials, permanent exhibition with the most important exhibits. Furthermore, the results of yearly polls of our visitors and comparison with relevant Euro-barometer polls are presented. (authors)

The pilot program described has been initiated under the Department of Energy Light Water Reactor Safety Research and Development Program and has the following specific objectives: to identify the principal types and causes of failures in valves, valve operators and their controls and associated hardware, which lead to, or could lead to plant trip; and to suggest possible remedies for the prevention of these failures and recommend future research and development programs which could lead to minimizing these valve failures or mitigating their effect on plant operation. The data surveyed cover incidents reported over the six-year period, beginning 1973 through the end of 1978. Three sources of information on valve failures have been consulted: failure data centers, participating organizations in the nuclear power industry, and technical documents.

The presence of caretakers/comforters during nuclear medicine examinations is relatively common. These caretakers receive higher doses than the general public, who receive only environmental/background exposure. The aim of this research was to know about the doses received by two significant groups of caretakers: comforters of cancer patients (Group I) and mothers of small children (Group II). The patients were scheduled to undergo two different diagnostic studies: Inmuno-Scintigraphy using a monoclonal antibody bound to (99m)Tc (for adults) and Renal Scintigraphy using (99m)Tc-dimercaptosuccinic acid (for children). The average effective doses were 0.27 and 0.29 mSv for Groups I and II, respectively. Additionally, environmental monitoring was performed in the waiting room for injected patients (Room I) and inside the procedure room (Room II). Equivalent environmental doses of 0.28 and 0.24 mSv for Rooms 1 and II, respectively, were found, which are similar to values reported by other authors. PMID:22517979

The practice of nuclear medicine requires integration of radiation safety with patient care and radiopharmaceutical standards. Nationally there was useful discussion in the UK before the Ionising Radiations Regulations and Approved Code of Practice were published, although such consultation had been lacking when the Medicines Act was implemented. Most of the new considerations relating to nuclear medicine stem from Schedule 6 of the Regulations. Generally, the presence of a single patient does not require a controlled area. However, when several patients are present, or radiopharmaceuticals are being prepared prior to injection, a controlled area is required. Classification of workers is not likely to be required in a typical nuclear medicine department in the UK, although most parts of the nuclear medicine department will need to be controlled areas. These include the radiopharmacy, radionuclide dispensary, injection room, and imaging rooms if patients are injected in them. The importance of finger dose measurements is emphasised. Patient wards, however, need not be controlled areas. A particular concern in nuclear medicine was that patients should not need to be admitted to hospital merely to comply with legislation. This is possibly the case and clarification will probably be available when the Notes for Guidance are published. Most procedures in nuclear medicine departments will remain unchanged. Further information is required, however, on patient waiting rooms, handling flood sources, pregnancy, and breast feeding. Within the hospital, detailed and multidisciplinary discussion will need to take place within the forum of the radiation safety committee. PMID:3664186

To date, discussions of the many different types of potential applications of PRA/PSA results and insights to safety-decision-making have been mainly theoretical. Various safety goals have been proposed as decision criteria. However, the discussion on the role of PRA/PSA and Safety Goals in safety-decision-making, especially in licensing, is controversial. A Working Group of the OECD Nuclear Energy Agency is completing a compilation and evaluation of real examples of past and present practical experience with the application of probabilistic methods in reactor safety decision-making, with the idea of developing a common understanding in this area. More than fifty different cases where PRA has influenced decision-making have been surveyed. These include, for example, regulatory changes, fixing of licensing requirements, plant specific modifications of design of operation, prioritization of safety issues and emergency planning. This feedback of experience - both positive and negative - with PRA/PSA applications is considered to provide guidance on how probabilistic approaches can be introduced into current safety practices, and on desirable future developments in probabilistic methods and specific PSA/PRA studies. Generic insights from the survey are given.

This report documents an inventory of calcined waste produced at the Idaho Nuclear Technology and Engineering Center during the period from December 1963 to May 2000. The report was prepared based on calciner runs, operation of the calcined solids storage facilities, and miscellaneous operational information that establishes the range of chemical compositions of calcined waste stored at Idaho Nuclear Technology and Engineering Center. The report will be used to support obtaining permits for the calcined solids storage facilities, possible treatment of the calcined waste at the Idaho National Engineering and Environmental Laboratory, and to ship the waste to an off-site facility including a geologic repository. The information in this report was compiled from calciner operating data, waste solution analyses and volumes calcined, calciner operating schedules, calcine temperature monitoring records, and facility design of the calcined solids storage facilities. A compact disk copy of this report is provided to facilitate future data manipulations and analysis.

MDR relays manufactured by Potter & Brumfield (P&B) have been used in various safety related applications in commercial nuclear power plants. These include emergency safety features (ESF) actuation systems, emergency core cooling systems (ECCS) actuation, and reactor protection systems. The MDR relays manufactured prior to May 1990 showed signs of generic failure due to corrosion and outgassing of coil varnish. P&B has made design changes to correct these problems in relays manufactured after May 1990. However, P&B does not manufacture the relays under any 10CFR50 Appendix B quality assurance (QA) program. They manufacture the relays under their commercial QA program and supply these as commercial grade items. This necessitates CGI Dedication of these relays for use in nuclear-safety-related applications. This paper presents a CGI dedication program that has been used to dedicate the MDR relays manufactured after been used to dedicate the MDR relays manufactured after May 1990. The program is in compliance with current Nuclear Regulatory Commission (NRC) and Electric Power Research Institute (EPRI) guidelines and applicable industry standards; it specifies the critical characteristics of the relays, provides the tests and analysis required to verify the critical characteristics, the acceptance criteria for the test results, performs source verification to quality P&B for its control of the critical characteristics, and provides documentation. The program provides reasonable assurance that the new MDR relays will perform their intended safety functions.

The Independent Safety Assessment described in this study report was performed to assess the safety of the design and launch plans anticipated by the U.S. Department of Defense (DOD) in 1993 for a Russian-built, U.S.-modified, TOPAZ-II space nuclear reactor power system. Its conclusions, and the bases for them, were intended to provide guidance for the U.S. Department of Energy (DOE) management in the event that the DOD requested authorization under section 91b. of the Atomic Energy Act of 1954, as amended, for possession and use (including ground testing and launch) of a nuclear-fueled, modified TOPAZ-II. The scientists and engineers who were engaged to perform this assessment are nationally-known nuclearsafety experts in various disciplines. They met with participants in the TOPAZ-II program during the spring and summer of 1993 and produced a report based on their analysis of the proposed TOPAZ-II mission. Their conclusions were confined to the potential impact on public safety and did not include budgetary, reliability, or risk-benefit analyses.

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the NuclearSafety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE NuclearSafety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.

This paper describes research aimed at improving regulatory supervision of radiation safety during work associated with the management of spent nuclear fuel and radioactive waste at legacy sites in northwest Russia through timely identification of employees presenting unfavourable human factor risk characteristics. The legacy sites of interest include sites of temporary storage now operated by SevRAO on behalf of Rosatom. The sites were previously operational bases for servicing nuclear powered submarines and are now subject to major remediation activities. These activities include hazardous operations for recovery of spent nuclear fuel and radioactive waste from sub-optimal storage conditions. The paper describes the results of analysis of methods, procedures, techniques and informational issues leading to the development of an expert-diagnostic information system for monitoring of workers involved in carrying out the most hazardous operations. The system serves as a tool for human factor and professional reliability risk monitoring and has been tested in practical working environments and implemented as part of regulatory supervision. The work has been carried out by the Burnasyan Federal Medical Biophysical Center, within the framework of the regulatory cooperation programme between the Federal Medical-Biological Agency of Russia and the Norwegian Radiation Protection Authority. PMID:23186692

XUNDL was established in 1998. While it contains compiled nuclear structure data in the "ENSDF" format, the difference is that the ENSDF collection holds evaluated data, while XUNDL holds experimental data that have not yet been evaluated. The manual for one of the evaluated collections defines evaluation as "the process of analyzing experimentally measured cross-section data, combining them with the predictions of nuclear model calculations, and attempting to extract the true value of a cross section." (Document ENDF-102) The XUNDL data are the experimental data before these calculations and comparisons are done. In general, the information in a given XUNDL dataset comes from a single journal article or from a set of closely related articles by one group of authors. The bulk of the compilations are carried out by the Nuclear Data Group at McMaster University. As of February 2008, the XUNDL database contained more than 2500 datasets for 1484 nuclides. The data can be searched through specialized interfaces or browsed. There is also a special list of hypernucleus datasets available. (Specialized Interface)

The objectives of this meeting were to capture the observations, insights, issues, concerns, and ideas of those involved in the Texas A&M University Nuclear Science Center (TAMU NSC) TRIGA Reactor Conversion so that future efforts can be conducted with greater effectiveness, efficiency, and with fewer challenges. This workshop was held in conjunction with a similar workshop for the University of Florida Reactor Conversion. Some of the generic lessons from that workshop are included in this report for completeness.

McClellan's Nuclear Radiation Center (MNRC) control console is in the process of being replaced due to spurious scrams, outdated software, and obsolete parts. The intent of the new control console is to eliminate the existing problems by installing a UNIX-based computer system with industry-standard interface software and incorporating human factors during all stages of the graphical user interface (GUI) development and control console design.

The use of peripherally inserted central catheter (PICC) line for central venous access in thermally injured patients has increased in recent years despite a lack of evidence regarding safety in this patient population. A recent survey of invasive catheter practices among 44 burn centers in the United States found that 37% of burn units use PICC lines as part of their treatment protocol. The goal of this study was to compare PICC-associated complication rates with the existing literature in both the critical care and burn settings. The methodology involved is a single institution retrospective cohort review of patients who received a PICC line during admission to a regional burn unit between 2008 and 2013. Fifty-three patients were identified with a total of seventy-three PICC lines. The primary outcome measurement for this study was indication for PICC line discontinuation. The most common reason for PICC line discontinuation was that the line was no longer indicated (45.2%). Four cases of symptomatic upper extremity deep vein thrombosis (5.5%) and three cases of central line-associated bloodstream infection (4.3%, 2.72 infections per 1000 line days) were identified. PICC lines were in situ an average of 15 days (range 1 to 49 days). We suggest that PICC line-associated complication rates are similar to those published in the critical care literature. Though these rates are higher than those published in the burn literature, they are similar to central venous catheter-associated complication rates. While PICC lines can be a useful resource in the treatment of the thermally injured patient, they are associated with significant and potentially fatal risks. PMID:25501778

The accomplishments during the months of July and August in the research and development program under way at ORNL as part of the U.S. Atomic Energy Commission's NuclearSafety Program are summarized, Included in this report are work on various chemical reactions, as well as the release, characterization, and transport of fission products in containment systems under various accident conditions and on problems associated with the removal of these fission products from gas streams. Although most of this work is in general support of water-cooled power reactor technology, including LOFT and CSE programs, the work reflects the current safety problems, such as measurements of the prompt fuel element failure phenomena and the efficacy of containment spray and pool-suppression systems for fission-product removal. Several projects are also conducted in support of the high-temperature gas-cooled reactor (HTGR). Other major projects include fuel-transport safety investigations, a series of discussion papers on various aspects of water-reactor technology, antiseismic design of nuclear facilities, and studies of primary piping and steel, pressure-vessel technology. Experimental work relative to pressure-vessel technology includes investigations of the attachment of nozzles to shells and the implementation of joint AEX-PVFX programs on heavy-section steel technology and nuclear piping, pumps, and valves. Several of the projects are directly related to another major undertaking; namely, the AEC's standards program, which entails development of engineering safeguards and the establishment of codes and standards for government-owned or -sponsored reactor facilities. Another task, CHORD-S, is concerned with the establishment of computer programs for the evaluation of reactor design data, The recent activities of the NSIC and the NuclearSafety journal in behalf of the nuclear community are also discussed.

The annual Research Program Plan describes publicly funded nuclear energy related research to be carried out mainly at the Technical Research Center of Finland (VTT) in 1992. The research is financed primarily by the Ministry of Trade and Industry (KTM), the Finnish Center for Radiation and NuclearSafety (STUK), and VTT itself. Other research institutes, utilities, and industry also contribute to many projects.

The objective of this study was to establish the initial quantitative reliability bounds for nuclear electric propulsion systems in a manned Mars mission required to ensure crew safety and mission success. Finding the reliability bounds involves balancing top-down (mission driven) requirements and bottom-up (technology driven) capabilities. In seeking this balance we hope to accomplish the following: (1) provide design insights into the achievability of the baseline design in terms of reliability requirements, given the existing technology base; (2) suggest alternative design approaches which might enhance reliability and crew safety; and (3) indicate what technology areas require significant research and development to achieve the reliability objectives.

Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

Five years ago the dramatic events in Fukushima that followed the massive earthquake and subsequent tsunami that struck Japan on March 11, 2011 sharpened the focus of scientists, engineers and general public on the broad range of technical, environmental and societal issues involved in assuring the safety of the world's nuclear power complex. They also called into question the potential of nuclear power to provide a growing, sustainable resource of CO2-free energy. The issues raised by Fukushima Dai-ichi have provoked urgent concern, not only because of the potential harm that could result from severe accidents or from intentional damage to nuclear reactors or to facilities involved in the nuclear fuel cycle, but also because of the extensive economic impact of those accidents and of the measures taken to avoid them.

The Great Tohoku Earthquake and Tsunami, and the severe accident at Fukushima Dai-ichi nuclear power plant 2011 alerted the nuclear industry to danger of extreme rare natural hazards. The subsequent "stress tests" performed by the nuclear industry in Europe and all over the world identifies the nuclear power plant (NPP) vulnerabilities and define the measures for increasing the plant safety. According to the international practice of nuclearsafety regulations, the cumulative core damage frequency for NPPs has to be 10-5/a, and the cumulative frequency of early large release has to be 10-6/a. In case of operating plants these annual probabilities can be little higher, but the licensees are obliged to implement all reasonable practicable measures for increasing the plant safety. For achieving the required level of safety, design basis of NPPs for natural hazards has to be defined at the 10-4/a ⎯10-5/a levels of annual exceedance probability. Tornado hazard is some kind of exception, e.g., the design basis annual probability for tornado in the US is equal to 10-7/a. Design of the NPPs shall provide for an adequate margin to protect items ultimately necessary to prevent large or early radioactive releases in the event of levels of natural hazards exceeding those to be considered for design. The plant safety has to be reviewed for accounting the changes of the environmental conditions and natural hazards in case of necessity, but as minimum every ten years in the frame of periodic safety reviews. Long-term forecast of environmental conditions and hazards has to be accounted for in the design basis of the new plants. Changes in hydroclimatic variables, e.g., storms, tornadoes, river floods, flash floods, extreme temperatures, droughts affect the operability and efficiency as well as the safety the NPPs. Low flow rates and high water temperature in the rivers may force to operate at reduced power level or shutdown the plant (Cernavoda NPP, Romania, August 2009). The

Enhancing the polarization of nuclear spins surrounding the nitrogen-vacancy (NV) center in diamond has recently attracted widespread attention due to its various applications. Here we present an analytical formula that not only provides a clear physical picture for the recently observed polarization reversal of strongly coupled13C nuclei over a narrow range of magnetic field [H. J. Wang et al., Nat. Commun. 4, 1940 (2013)], but also demonstrates the possibility to strongly polarize weakly coupled13C nuclei. This allows sensitive magnetic field control of the 13C nuclear spin polarization for NMR applications and significant suppression of the 13C nuclear spin noise to prolong the NV spin coherence time. PMID:26521962

The U.S. Department of Energy, Richland Operations Office (RL) is responsible for 30 hazard category 2 and 3 nuclear facilities that are operated by its prime contractors, Fluor Hanford Incorporated (FHI), Bechtel Hanford, Incorporated (BHI) and Pacific Northwest National Laboratory (PNNL). The publication of Title 10, Code of Federal Regulations, Part 830, Subpart B, Safety Basis Requirements (the Rule) in January 2001 imposed the requirement that the Documented Safety Analyses (DSA) for these facilities be reviewed against the requirements of the Rule. Those DSA that do not meet the requirements must either be upgraded to satisfy the Rule, or an exemption must be obtained. RL and its prime contractors have developed a NuclearSafety Strategy that provides a comprehensive approach for supporting RL's efforts to meet its long term objectives for hazard category 2 and 3 facilities while also meeting the requirements of the Rule. This approach will result in a reduction of the total number of safety basis documents that must be developed and maintained to support the remaining mission and closure of the Hanford Site and ensure that the documentation that must be developed will support: compliance with the Rule; a ''Just-In-Time'' approach to development of Rule-compliant safety bases supported by temporary exemptions; and consolidation of safety basis documents that support multiple facilities with a common mission (e.g. decontamination, decommissioning and demolition [DD&D], waste management, surveillance and maintenance). This strategy provides a clear path to transition the safety bases for the various Hanford facilities from support of operation and stabilization missions through DD&D to accelerate closure. This ''Just-In-Time'' Strategy can also be tailored for other DOE Sites, creating the potential for large cost savings and schedule reductions throughout the DOE complex.

The Department of Energy (Department) submits an Annual Report to Congress each year detailing the Department’s activities relating to the Defense Nuclear Facilities Safety Board (Board), which provides advice and recommendations to the Secretary of Energy (Secretary) regarding public health and safety issues at the Department’s defense nuclear facilities. In 2003, the Department continued ongoing activities to resolve issues identified by the Board in formal recommendations and correspondence, staff issue reports pertaining to Department facilities, and public meetings and briefings. Additionally, the Department is implementing several key safety initiatives to address and prevent safety issues: safety culture and review of the Columbia accident investigation; risk reduction through stabilization of excess nuclear materials; the Facility Representative Program; independent oversight and performance assurance; the Federal Technical Capability Program (FTCP); executive safety initiatives; and quality assurance activities. The following summarizes the key activities addressed in this Annual Report.

Patient-centered technologies have emerged as a way to actively engage patients in care. The reach and potential of cell phones to engage diverse patient populations is great. Evidence of their effectiveness in improving health-related outcomes is limited. Researchers conducted an online survey of community health centers and clinics to assess if and how health care providers in the safety net use cell phones to support patient engagement. The findings indicate that the use of cell phones in patient care is at an early stage of deployment across the safety net. Organizations identify chronic disease management as an area where cell phones offer considerable potential to effectively engage patients. To promote widespread adoption and use, technical assistance to support the implementation and management of interventions, evidence-based or best practice models that highlight successful implementation strategies in care delivery, and the introduction of new payment or reimbursement policies will be essential. PMID:26040018

... Relinquishment From Ryder Trauma Center AGENCY: Agency for Healthcare Research and Quality (AHRQ), HHS. ACTION... accepted a notification of voluntary relinquishment from Ryder Trauma Center of its status as a PSO, and... PSOs. AHRQ has accepted a notification from Ryder Trauma Center, PSO number P0019, which is a...

This formal monthly report covers the studies related to the use of /sup 238/PuO/sub 2/ in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of the Los Alamos Scientific Laboratory. The two programs involved are: General-Purpose Heat Source Development and Space NuclearSafety and Fuels. Most of the studies discussed here are of a continuing nature. Results and conclusions described may change as the work continues. Published reference to the results cited in this report should not be made without the explicit permission of the person in charge of the work.

This paper presents a logical approach to designing a safety test plan for a space nuclear system. It is pointed out that two important facts need to underlie the development of a test plan: first, that sequential insults and the accumulation of damage are the rule; and second that the response of the nuclear system is stochastic (i.e., for any given set of conditions a probabilistic range of outcomes will occur regardless of the state of our knowledge). Because of these facts a deterministic approach can only be a starting point. The substance of the approach consists of undertaking and documenting three basic efforts: (1) a description of the analysts view of the problem and how it fits into the safety analysis, (2) a formal documentation of the purpose and requirements of the test plan (or test), and (3) an assessment of the use or usefulness of existing test data.

This paper presents a logical approach to designing a safety test plan for a space nuclear system. It is pointed out that two important facts need to underlie the development of a test plan: first, that sequential insults and the accumulation of damage are the rule; and second that the response of the nuclear system is stochastic (i.e., for any given set of conditions a probabilistic range of outcomes will occur regardless of the state of our knowledge). Because of these facts a deterministic approach can only be a starting point. The substance of the approach consists of undertaking and documenting three basic efforts: (1) a description of the analysts view of the problem and how it fits into the safety analysis, (2) a formal documentation of the purpose and requirements of the test plan (or test), and (3) an assessment of the use or usefulness of existing test data.

The paper first briefly reviews the neutron source multiplication method and then presents an experimental study that shows that the parameter measured by the neutron source multiplication method actually is a subcritical effective neutron multiplication factor k{sub s} with an external neutron source, not the effective neutron multiplication factor k{sub eff}. The parameters k{sub s} and k{sub eff} have been researched for a nuclear critical safety experiment assembly using a uranium solution. The parameter k{sub s} was measured by the source multiplication method, while the parameter k{sub eff} was measured by the power-raising period method. The relationship between k{sub eff} and k{sub s} is discussed and their effects on nuclearsafety are mentioned.

The Seismic Safety Margins Research Program (SSMRP) has a gola to develop a complete fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-induced radioactive release from a commercial nuclear power plant. As part of this program, calculations of the seismic risk from a typical commercial nuclear reactor were made. These calculations required a knowledge of the probability of failure (fragility) of safety-related components in the reactor system which actively participate in the hypothesized accident scenarios. This report describes the development of the required fragility relations and the data sources and data reduction techniques upon which they are based. Both building and component fragilities are covered. The building fragilities are for the Zion Unit 1 reactor which was the specific plant used for development of methodology in the program. Some of the component fragilities are site-specific also, but most would be usable for other sites as well.

Current robotic systems have been developed primarily for the automotive and electronic industry. The adaptation of these commercial robotic systems to applications in the manufacturing of nuclear fuel requires the addition of safety interlocks as to the handling and accountability of nuclear materials. Also, additional safety interlocks are required when the robots are operated in containment enclosures that are environmentally sealed. Interlocks have been incorporated into a commercial robot. The robotic system has been installed in the containment enclosure as part of the pellet storage subsystem into the Secure Automated Fabrication (SAF) facility currently being built by Westinghouse Hanford Company (WHC) for the US Department of Energy (DOE). The system has been installed in the Fuel Cycle Plant and is scheduled for initial operational testing in 1986.

We completed a novel assessment to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor (PWR) sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance.more » These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δkeff were observed; calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014); and significant uncredited transportation dose rate margins were also observed. The results demonstrate that, at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associated with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.« less

The Spent Nuclear Fuels Project (SNFP) is charged with moving to storage 2,100 metric tons of spent nuclear fuel elements left over from plutonium production at DOE'S Hanford site in Washington state. Two new facilities, the Cold Vacuum Drying Facility (CVDF) and the Canister Storage Building (CSB) are in final construction. In order to meet aggressive schedule commitments, the SNFP chose to prepare the safety analysis reports (SAR's) in phases that covered only specific portions of each facility's design as it was built. Each SAR also merged the preliminary and final safety analysis reports into a single SAR, thereby covering all aspects of design, construction, and operation for that portion (phase) of the facility. A policy of ''NRC equivalency'' was also implemented in parallel with this effort, with the goal of achieving a rigor of safety analysis equivalent to that of NRC-licensed fuel processing facilities. DOE Order 5480.23. ''NuclearSafety Analysis Reports'' allows preparation of both a phased and a merged SAR to accelerate construction schedules. However, project managers must be aware that such acceleration is not guaranteed. Managers considering this approach for their project should be cognizant of numerous obstacles that will be encountered. Merging and phasing SAR's will create new, unique, and unanticipated difficulties which may actually slow construction unless expeditiously and correctly managed. Pitfalls to be avoided and good practices to be implemented in preparing phased and merged SAR's are presented. The value of applying NRC requirements to the DOE safety analysis process is also discussed. As of December, 1999, the SNFP has completed and approved a SAR for the CVDF. Approval of the SAR for the CSB is pending.

... technical framework and requirements for interoperability in order to ensure that the public safety wireless broadband network is interoperable on a nationwide basis. In particular, the Committee will provide... broadband networks, and other matters related to the functioning of the nationwide public safety...

The paper will focus on the discussion on implementation of external hazards in the probabilistic safety assessment (PSA) methods for the extreme external hazards mainly focused on Seismic, Flooding, Meteorological Hazards (e.g. Storm, Extreme temperature, snow pack), Biological infestation, Lightening hazards, Accidental Aircraft crash and man- made hazards including natural external fire and external explosion. This will include discussion on identification of some good practices on the implementation of external hazards in Level 1 PSA, with a perspective of development of extended PSA and introduction of relevant modelling for external hazards in an existing Level 1 PSA. This paper is associated to the European project ASAMPSAE (www.asampsa.eu) which gathers more than 30 organizations (industry, research, safety control) from Europe, US and Japan and which aims at identifying some meaningful practices to extend the scope and the quality of the existing probabilistic safety analysis developed for nuclear power plants.

The Hanford Nuclear Reservation is one of the U.S. Department of Energy's largest nuclear weapons sites. The enormous changes experienced by Hanford over the last several years, as its mission has shifted from weapons production to cleanup, has profoundly affected its occupational health and safety services. Innovative programs and new initiatives hold promise for a safer workplace for the thousands of workers at Hanford and other DOE sites. However, occupational health and safety professionals continue to face multiple organizational, economic, and cultural challenges. A major problem identified during this review was the lack of coordination of onsite services. Because each health and safety program operates independently (albeit with the guidance of the Richland field operations office), many services are duplicative and the health and safety system is fragmented. The fragmentation is compounded by the lack of centralized data repositories for demographic and exposure data. Innovative measures such as a questionnaire-driven Employee Job Task Analysis linked to medical examinations has allowed the site to move from the inefficient and potentially dangerous administrative medical monitoring assignment to defensible risk-based assignments and could serve as a framework for improving centralized data management and service delivery. PMID:11186038

Nuclear power reactor safety in the United States is about to enter a new era -- an era of risk- based management and risk-based regulation. First, there was the age of `prescribed safety assessment,' during which a series of design-basis accidents in eight categories of severity, or classes, were postulated and analyzed. Toward the end of that era, it was recognized that `Class 9,' or `beyond design basis,' accidents would need special attention because of the potentially severe health and financial consequences of these accidents. The accident at Three Mile Island showed that sequences of low-consequence, high-frequency events and human errors can be much more risk dominant than the Class 9 accidents. A different form of safety assessment, PSA, emerged and began to gain ground against the deterministic safety establishment. Eventually, this led to the current regulatory requirements for individual plant examinations (IPEs). The IPEs can serve as a basis for risk-based regulation and management, a concept that may ultimately transform the U.S. regulatory process from its traditional deterministic foundations to a process predicated upon PSA. Beyond the possibility of a regulatory environment predicated upon PSA lies the possibility of using PSA as the foundation for managing daily nuclear power plant operations.

Following the hydraulic rupture of a UF cylinder at the Sequoyah Fuels Facility on January 4, 1986, the US Nuclear Regulatory Commission's (NRC's) executive director for operations (EDO) established an augmented inspection team to investigate the accident. The investigation is reported in NUREG-1179. The EDO then formed a lessons-learned group to report on the action NRC might reasonably take to prevent similar accidents. The group's recommendations are reported in NUREG-1198. In addition, the EDO formed an independent materials safety regulation review study group (MSRRSG) to review the licensing and inspection program for NRC-licensed fuel cycle and materials facilities. During the same period of time that the MSRRSG report was being prepared and evaluated, the staff undertook an independent action to assess operational safety at each of the 12 major fuel facilities licensed by the NRC. The facilities included the 2 facilities producing uranium hexafluoride, the 7 facilities producing commercial nuclear reactor fuel, and the 3 facilities producing naval reactor fuel. The most important safety issues identified as needing attention by licensees were in the areas of fire protection, chemical hazards identification and mitigation, management controls or quality assurance, safety-related instrumentation and maintenance, and emergency preparedness.

The recent 6 October 1990 launch and deployment of the nuclear-powered Ulysses spacecraft from the Space Shuttle {ital Discovery} culminated an extensive safety review and evaluation effort by the Interagency NuclearSafety Review Panel (INSRP). After more than a year of detailed independent review, study, and analysis, the INSRP prepared a Safety Evaluation Report (SER) on the Ulysses mission, in accordance with Presidential Directive-National Security Council memorandum 25. The SER, which included a review of the Ulysses Final Safety Analysis Report (FSAR) and an independent characterization of the mission risks, was used by the National Aeronautics and Space Administration (NASA) in its decision to request launch approval as well as by the Executive Office of the President in arriving at a launch decision based on risk-benefit considerations. This paper provides an overview of the Ulysses mission and the conduct as well as the results of the INSRP evaluation. While the mission risk determined by the INSRP in the SER was higher than that characterized by the Ulysses project in the FSAR, both reports indicated that the radiological risks were relatively small. In the final analysis, the SER proved to be supportive of a positive launch decision. The INSRP evaluation process has demonstrated its effectiveness numerous times since the 1960s. In every case, it has provided the essential ingredients and perspective to permit an informed launch decision at the highest level of our Government.

In 2009, the International Safety Framework for Nuclear Power Source Applications in Outer Space was adopted, following a multi-year process that involved all major space faring nations under the auspices of a partnership between the UN Committee on the Peaceful Uses of Outer Space and the International Atomic Energy Agency. The Safety Framework reflects an international consensus on best practices to achieve safety. Following the 1992 UN Principles Relevant to the Use of Nuclear Power Sources in Outer Space, it is the second attempt by the international community to draft guidance promoting the safety of applications of nuclear power sources in space missions. NPS applications in space have unique safety considerations compared with terrestrial applications. Mission launch and outer space operational requirements impose size, mass and other space environment limitations not present for many terrestrial nuclear facilities. Potential accident conditions could expose nuclear power sources to extreme physical conditions. The Safety Framework is structured to provide guidance for both the programmatic and technical aspects of safety. In addition to sections containing specific guidance for governments and for management, it contains technical guidance pertinent to the design, development and all mission phases of space NPS applications. All sections of the Safety Framework contain elements directly relevant to engineers and space mission designers for missions involving space nuclear power sources. The challenge for organisations and engineers involved in the design and development processes of space nuclear power sources and applications is to implement the guidance provided in the Safety Framework by integrating it into the existing standard space mission infrastructure of design, development and operational requirements, practices and processes. This adds complexity to the standard space mission and launch approval processes. The Safety Framework is deliberately

Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for considerably longer periods than originally intended (e.g., <40 years). Extended storage (ES) time and irradiation of nuclear fuel to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. This effort is primarily motivated by concerns related to the potential for fuel degradation during ES periods and transportation following ES. The criticality analyses consider representative UNF designs and cask systems and a range of fuel enrichments, burnups, and cooling times. The various failed-fuel configurations considered are designed to bound the anticipated effects of individual rod and general cladding failure, fuel rod deformation, loss of neutron absorber materials, degradation of canister internals, and gross assembly failure. The results quantify the potential impact on criticality safety associated with fuel reconfiguration and may be used to guide future research, design, and regulatory activities. Although it can be concluded that the criticality safety impacts of fuel reconfiguration during transportation subsequent to ES are manageable, the results indicate that certain configurations can result in a large increase in the effective neutron multiplication factor, k{sub eff}. Future work to inform decision making relative to which configurations are credible, and therefore need to be considered in a safety evaluation, is recommended.

A common problem in ethics is that people often desire an end but fail to take the means necessary to achieve it. Employers and employees may desire the safety end mandated by performance standards for pollution control, but they may fail to employ the means, specification standards, necessary to achieve this end. This article argues that current (de jure) performance standards, for lowering employee exposures to ionizing radiation, fail to promote de facto worker welfare, in part because employers and employees do not follow the necessary means (practices known as specification standards) to achieve the end (performance standards) of workplace safety. To support this conclusion, the article argues that (1) safety requires attention to specification, as well as performance, standards; (2) coal-mine specification standards may fail to promote performance standards; (3) nuclear workplace standards may do the same; (4) choosing appropriate means to the end of safety requires attention to the ways uncertainties and variations in exposure may mask violations of standards; and (5) correcting regulatory inattention to differences between de jure and de facto is necessary for achievement of ethical goals for safety. PMID:15028007

Bogazici University and Kandilli Observatory and Earthquake Research Institute (KOERI) is acting as the Turkish National Data Center (NDC) and responsible for the operation of the International Monitoring System (IMS) Primary Seismic Station (PS-43) under Belbasi Nuclear Tests Monitoring Center for the verification of compliance with the Comprehensive Nuclear-Test-Ban Treaty (CTBT) since February 2000. The NDC is responsible for operating two arrays which are part of the IMS, as well as for transmitting data from these stations to the International Data Centre (IDC) in Vienna. The Belbasi array was established in 1951, as a four-element (Benioff 1051) seismic array as part of the United States Atomic Energy Detection System (USAEDS). Turkish General Staff (TGS) and U.S. Air Force Technical Application Center (AFTAC) under the Defense and Economic Cooperation Agreement (DECA) jointly operated this short period array. The station was upgraded and several seismometers were added to array during 1951 and 1994 and the station code was changed from BSRS (Belbasi Seismic Research Station) to BRTR-PS43 later on. PS-43 is composed of two sub-arrays (Ankara and Keskin): the medium-period array with a ~40 km radius located in Ankara and the short-period array with a ~3 km radius located in Keskin. Each array has a broadband element located at the middle of the circular geometry. Short period instruments are installed at depth 30 meters from the surface while medium and broadband instruments are installed at depth 60 meters from surface. On 25 May 2009, The Democratic People’s Republic of Korea (DPRK) claimed that it had conducted a nuclear test. Corresponding seismic event was recorded by IMS and IDC released first automatic estimation of time (00:54:43 GMT), location (41.2896°N and 129.0480°E) and the magnitude (4.52 mb) of the event in less than two hours time (USGS: 00:54:43 GMT; 41.306°N, 129.029°E; 4.7 mb) During our preliminary analysis of the 25th May 2009 DPRK

It has been suggested that the central regions of many galaxies are unlikely to be in a static steady state, with instabilities caused by sinking satellites, the influence of a supermassive black hole or residuals of galaxy formation, resulting in the nuclear black hole orbiting the galaxy center. The observational signature of such an orbiting black hole is an offset of the active nucleus (AGN) from the kinematic center defined by the galaxy rotation curve. This orbital motion may provide fuel for the AGN, as the hole 'grazes' on the ISM, and bent radio jets, due to the motion of their source. The early type (E/SO) Seyfert galaxy, NGC2210, with its striking twin, 'S'-shaped radio jets, is a unique and valuable test case for the offset-nucleus phenomenon since, despite its remarkably normal rotation curve, its kinematically-measured mass center is displaced both spatially (260 pc) and kinematically (170 km/s) from the active nucleus located in optical and radio studies. However, the central kinematics, where the rotation curve rises most steeply, have been inaccessible with ground-based resolutions. We present new, high resolution WFPC2 imaging and long-slit STIS spectroscopy of the central 300 pc of NGC2110. We discuss the structure and kinematics of gas moving in the galactic potential on subarcsecond scales and the reality of the offset between the black hole and the galaxy mass center.

The F center in alkali halides is a well-known prototype of a strongly coupled localized electron-phonon system. This colour center is one of the long studied targets in the field of photophysics because it is simple but rich in variety. Steady-state spectroscopy, such as modulation spectroscopy and Raman scattering spectroscopy, has elucidated the strength of the electron-phonon coupling in the (meta-)stable state, i.e. the ground state and the relaxed excited state. Picosecond spectroscopy has improved understanding of the state mixing in the transient state. Owing to recent developments of ultrafast lasers with pulse widths shorter than oscillation periods of phonons, it has been possible to perform real-time observation of lattice vibration, and the understanding of the transient state has been remarkably expanded. In this paper, we review early and present studies on dynamics of electron-phonon coupling at the F center, especially recent real-time observations on the dynamics of nuclear wave packets in the excited state of the F center in KI, KBr, KCl and RbCl. These real-time observations reveal (i) spatial extension of the electronic wave function of a trapped electron, (ii) the difference between the coupled phonons in the ground state and the excited state, (iii) diabatic transition between the adiabatic potential energy surfaces and (iv) anharmonicity of the potential energy surface.

A research, development, and demonstration project for the grouting of newly generated liquid waste (NGLW) at the Idaho Nuclear Technology and Engineering Center is considered feasible. NGLW is expected from process equipment waste, decontamination waste, analytical laboratory waste, fuel storage basin waste water, and high-level liquid waste evaporator condensate. The potential grouted waste would be classed as mixed low-level waste, stabilized and immobilized to meet RCRA LDR disposal in a grouting process in the CPP-604 facility, and then transported to the state.

We propose a scheme enabling controlled quantum coherent interactions between separated nitrogen-vacancy centers in diamond in the presence of strong magnetic fluctuations. The proposed scheme couples nuclear qubits employing the magnetic dipole-dipole interaction between the electron spins and, crucially, benefits from the suppression of the effect of environmental magnetic field fluctuations thanks to a strong microwave driving. This scheme provides a basic building block for a full-scale quantum-information processor or quantum simulator based on solid-state technology. PMID:22107276

The purpose of the Workshop on Compositional and Microstructural Analysis of Nuclear Waste Materials, conducted November 11 and 12, 1980, was to critically examine and evaluate the various methods currently used to study non-radioactive, simulated, nuclear waste-form performance. Workshop participants recognized that most of the Materials Characterization Center (MCC) test data for inclusion in the Nuclear Waste Materials Handbook will result from application of appropriate analytical procedures to waste-package materials or to the products of performance tests. Therefore, the analytical methods must be reliable and of known accuracy and precision, and results must be directly comparable with those from other laboratories and from other nuclear waste materials. The 41 participants representing 18 laboratories in the United States and Canada were organized into three working groups: Analysis of Liquids and Solutions, Quantitative Analysis of Solids, and Phase and Microstructure Analysis. Each group identified the analytical methods favored by their respective laboratories, discussed areas needing attention, listed standards and reference materials currently used, and recommended means of verifying interlaboratory comparability of data. The major conclusions from this workshop are presented.

...On May 8, 2012, the Defense Nuclear Facilities Safety Board submitted Recommendation 2012-1, concerning Savannah River Site Building 235-F Safety, to the Department of Energy. In accordance with section 315(b) of the Atomic Energy Act of 1954, as amended, 42 U.S.C. 2286d(b), the following represents the Secretary of Energy's response to the...

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The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category.

The history of the NASA Lewis Research Center's role in space nuclear power programs is reviewed. Lewis has provided leadership in research, development, and the advancement of space power and propulsion systems. Lewis' pioneering efforts in nuclear reactor technology, shielding, high temperature materials, fluid dynamics, heat transfer, mechanical and direct energy conversion, high-energy propellants, electric propulsion and high performance rocket fuels and nozzles have led to significant technical and management roles in many natural space nuclear power and propulsion programs.